ML20202B106

Nuclear Safety Review Staff (Nsrs) Followup Review of Open Items from Previous Nsrs Reviews & Investigations

ML20202B106
Person / Time
Site:	Sequoyah
Issue date:	03/21/1986
From:	Bennett H, Debbage A, Oblock V TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20202B080	List: ML20202B075 ML20202B106
References
R-86-01-SQN, R-86-1-SQN, NUDOCS 8604110131
Download: ML20202B106 (105)

Text

• 001 '8s 0321 051 TENNESSEE VALLEY AUTHORITY a .

NUCLEAR SAFETY REVIEW STAFF NSRS REPORT NO. R-86-01-SQN

SUBJECT:

SEQUOYAH NUCLEAR PLANT (SQN) - NUCLEAR SAFETY REVIEW STAFF (NSRS) FOLLOW-UP REVIEW OF OPEN ITEMS FROM PREVIOUS NSRS REVIEWS AND INVESTIGATIONS.

DATES OF REVIEW: FEBRUAKY 3-25, 1986 REVIEWERS: 8'J/- [

V.S. O' BLOCK DATE M -

3 -Z/-8 6 H. W. BENNETT DATE 1/ 3 - A A - fl A. G. DEBBAGE DATE

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K. G. LAWLESS * '/ DATE APPROVED BY: 7/2//P-[

R.D.S % DATE /

• February 3 and 4, 1986, only.

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8604110131 860404 PDR ADOCK 05000327 l P PDR l

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TABLE OF CONTENTS PAGE I. SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 II. CONCLUSIONS AND RECOMMENDATIONS . . . . . . . . . . . . . . . . 1 III. STATUS OF OPEN ITEMS . . . . . . . . . . . . . . . . . . . . . . 3 IV. DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 V. LIST OF PERSONNEL CONTACTED . . . . . . . . . . . . . . . . . . 82 i

VI. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . 85 i

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This follow-up review was conducted to assess the status of open items related to the Sequoyah Nuclear Plant (SQN) that had been identified during ten previous Nuclear Safety Review Staff (NSRS) investigations and reviews. Two of these items (R-82-04-NPS-1 and R-85-02-SQN/WBN-01) were also reviewed for status at the Watts Bar Nuclear (WBN) plant. The status of the following 44 open items was reviewed:

o R-80-03-NUC PR-C1, C2 (see reference A.1) o R-80-05-SQN-4 (see reference A.2) o R-81-07-SQN-7 (see reference A.4) o R-82-04-NPS-1 (see reference A.6) o I-82-20-SQN-1, 2 (see reference A.7) o R-82-21-SQN-2, 4, 5, 6, 7, 9 (see reference A.11) o I-84-12-SQN- 1 through 23 (see reference A.25) o R-84-17-NPS-2 (see reference A.41) o R-85-02-SQN/WBN-1, 2 (see reference A.42) o R-85-03-NPS-1, 4, 6, 7, 8 (see reference A.54)

This follow-up review consisted of discussions with the Office of Nuclear i Power (OMP) personnel and evaluation of TVA documentation and corrective

action associated with each open item. i II. CONCLUSIONS AND REC 0letENDATIONS During this follow-up review, the status of 44 open items related to SQN from 10 NSRS review and investigation reports was reviewed. The NSRS considers 40 items closed and 4 remaining open with 1 additional recommendation generated. Of the four open items, the NSRS considers it j necessary that the corrective action associated with the two

recommendations R-80-05-SQN-4B and R-85-02-SQN/WBN-02 be completed prior to restart. The other two open items (R-85-03-NPS-07, -08) and the new reconunendation (R-86-01-SQN-01) are not deemed necessary to be resolved prior to restart; however, corrective action should be expedited. The q

status and details of these four open items is provided in sections III

and IV, respectively. The additional reconumendation follows.

1 A. R-86-01-SQN-01 Improvements in Overall As Low As Reasonably Achievable (ALARA) Program conclusion t

During the follow-up review of the NSRS reconunendation I-84-12-SQN-13 (see section IV.Z), it was determined that there are weaknesses in i the overall SQN ALARA program in the following areas- ,

1. Inadequate staffing to support the ALARA engineering effort.

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2. Lack of a plant /LARA r > view connaittee' with responsibility for

overall coordination of the ALARA program.

3. Ineffective ALARA employee suggestion program.

4. Ineffective ALARA coordination between site functional organizations.

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Detailed knowledge of ALARA techniques for many individuals *

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responsible for ALARA implementation.

Recommendation NSRS recommends that actions be taken to improve the effectiveness of the SQN ALARA program. The following suggestions should be used when determining what actions SQN will take:

1. ATADA Enmineer Staffinz 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, modifications, outage planning, design, and site services. A job-task analysis could be used to determine an effective staffing level.

2. ALARA Review Comunittee.. Establish an ALARA review connaittee composed of members from the major functional areas with the responsibility for overall coordination of the ALARA program.

Specific functions would include:

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a. Review exposure reduction plans for specific jobs with exposure estimates greater than 25 man-rem.

b. Direct the implementation of approved ALARA suggestions.

l c. Review planning schedules.

d. Review specific and timely ALARA problems; such as, reports of unnecessary loitering in dose areas.

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e. Review personnel contamination reports.

f. Review corrective action on delinquent postjob ALARA reports.

g. Review status of ALARA projects.

h. other.

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

3. ALARA Employee Summestion Proaram. Increase employee participation in the ALARA employee suggestion program.

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

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4. Department ALARA Coordinators. ALARA coordinators should be assigned to all site functional organizations, e.g.,

modifications, operations, maintenance, test, design, and site 4 services, to provide these groups with the expertise necessary to support all aspects of the ALARA program. This would be an i expansion of the current plans of the HP Section to assign an

! M-3 HP to assist planners with ALARA.

5. Trainina. An ALARA training program should be prepared and i given to those individuals directly responsible for the ALARA i plant efforts, e.g., ALARA Committee members, department coordinators, plus those individuals responsible for preparation of ALARA preplans and postplans. The training program should be l extensive and incorporate as basic elements: the physics of

. radiation; fundamentals of radiation attenuation; types of

! radiation sources; review of industry experience; methods to I reduce exposure, e.g., changing test frequency or time of test, changing preventive maintenance frequency or time of maintenance, relocate components with high failure rates to lower radiation fields, and/or add perinanent shielding, flushing

, systems, etc.

See section IV.SS for details of this recommendation..

I i III. STATUS OF OPEN ITEMS A. R-80-03-NUC PR-C1, Additional Trainina for STA In the original review, NSRS reconumended that additional training be

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provided to SQN Shift Technical Advisors (STAS) to assure cognizance of existing requirements for reviewing shif t engineer's journals. By

? reviewing on-the-job training material and having discussions with the Reactor Engineering Unit Supervisor and an STA it was determined in this follow-up review that adequate requirements exist in the

! training material for review of night order books, journals, and logs and that the on-the-job training provides the STA with an adequate understanding of these review requirements. This item is closed (see j section IV.A for details).

! B. R-80-03-NUC PR-C2, Adeouste STA Trainina Records

In the original review, NSRS recomunended that SQN review training records to assure that adequate training records are available to demonstrate that training has been completed. For this follow-up ,

review, it was determined that the Power Operations Training Center (POTC) maintains the en-the-job-training records for the SQN STAS.

Seseral of these on-the-job training records were reviewed and determined to be adequate. Discussion with one of these STAS verified that the individual had completed the training and possessed a thorough knowledge of the requirements for STA los and journal review prior to and after assuming the STA shift. This item is closed (see section IV.B for details).

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l l C R-80-05-SQN-04, Electrical Deficiencies . ,

Part 4 - Four specific electrical deficiencies were found in the i

original review. Three of these were closed by a follow-up review the same year, the fourth was verified complete in this follow-up ,

review. This item is closed (see section IV.C, Part A, for details). j Part 3 - In the original review, NSES recommended revising the l l

configuration control program to require verification of plant l l

configuration once per refueling cycle. The configuration control i I

programs have improved considerably since 1980. NSRS believes that once a baseline configuration is established, the configuration l

} control programs will serve to raintain control over the l configuration such that verification once per refueling cycle will not be required. This item remains open pending verification of the

confihuration baseline for critical structures, systems, and components (CSSC) by completion of the following actions prior to restart (see section IV.C, Part B, for details)

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a. Completion of Phase I of the plan for conversion to configuration control drawings for those drawings previously j identified by the plant as necessary for CSSC configuration control.

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j b. Selection, by the plant, of a number of CSSC systems for

complete walkdown to verify that the actual configuration agrees with the CSSC configuration control drawings verified as part of 2 the Phase I effort.

I i c. Walkdown of the selected systems and correction of any i discrepancies found. If significant problems are found, additional systems should be selected for walkdown.

j Part C - In the original review, NSRS recommended that programs be j implemented or revised for all employees to emphasize the need for i and maintenance of configuration control. BSRS found in this follow-up review that employees are made aware of the need for configuration control and their responsibilities through normal orientation and training in instructions and through regular exposure to configuration control in the instructions they normally use.

Employee awareness of configuration control appears to be adequate.

j This item is closed (see section IV.C. Part C, for details).

i In summary, Parts A and C are closed and Part B remains open.

l D. R-81-07-SQN-07, Unreviewed T crary Alteration Control Forins i

Port 1 - In the original review, NSRS found discrepancies between the

division procedure manual and AI-1.9, " Control of Temporary

! Alterations and Use of the Temporary Alteration control Order." In i 'this follow-up review, ISES found that the discrepancies between

! AI-1.9 and the new upper-tier document, the WQAM, were resolved.

This item is closed (see section IV.D. Part 1, for details).

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• Part 2 - In the original review, NSRS found many CSSC temporary alterations that had been in effect for nearly a year and recommended that the status of outstanding unit 2 CSSC temporary alterations be reviewed prior to fuel loading. For this follow-up review, NSRS found no evidence that this had been done. However, PORC review of all outstanding temporary alterations is now required every six months, monthly status reports are issued to managers and supervisors, a commitment was made to INPO to close specific old temporary alterations, and the number of outstanding temporary alterations have been steadily decreasing. This indicates an appropriate level of attention to the control of temporary alterations. This item is closed (see section IV.D. Part 2, for details).

E. R-82-04-NPS-01 Containment Spray Test Line at SON and WBN.

In the original review, NSRS recommended that automatic isolation of the containment spray test line at SQN and WBN be provided to enable isolation of this test line in the event an accident occurs requiring the use of the containment spray system when testing is in progress. For this follow-up review.it was determined that rather than automatic isolation being provided, the SQN containment spray pump test instruction SI-37, " Containment Spray Pump Test," was modified to require an Assistant Unit Operator (AUO) be stationed near the manual isolation valves during conduct of the pump testing and be in constant communication with the control room. Thus, if an accident requiring the containment spray system occurred during testing, the AUO would be instructed to close the test line valves.

This change was determined to be acceptable to NSRS. The WBN instruction SI-4.0.5.72-P, " Containment Spray Pump Test," has been revised in the same manner as the SQN instruction, namely, requiring an AU0 to be stationed at the test line isolation valves and be in contact with the control room during conduct of the test. This item is closed (see section IV.E for details).

F. I-82-20-SQN-01, Administration of KI to Plant Personnel As a result of the original investigation, NSRS determined that applicable procedures did not address the administration of KI uniformly and recommended that consistent guidance be provided. For this follow-up review, NSRS found that the appropriate documents have been revised to provide consistent guidance. This item is closed (see section IV.F for details).

G. I-82-20-SQN-02, Uptrade of Field Team Van As a result of the original investigation, NSRS recommended the addition of a permanent seat in the rear and compartmentalized labeled equipment storage for the field team vans. For this follow-up review, NSRS found that these recommendation have been implemented. This item is closed (see section IV.G for details).

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H. I-82-21-SQN-02 Emphasize Pre-Job plannina . .

As a result of the original investigation NSRS recomunended that additional emphasis should be placed on prejob planning and procedure development with review by both the Nuclear Central Office and SQN to ensure that hazards were identified and reduced to an acceptable '

level of risk and.that. proper equipment was available in working order prior to commencement of the activity. For this follow-up review, it was determined that new positions of maintenance planners have been created at SQN and are filled. Their job descriptions require them to do these necessary functions before the job is started which includes the removal of persons from situations,where hazards are not adequately controlled. This item is closed (see section IV.H for details).

I. I-82-21-SQN-04, Establish pronram to Evaluate Unusual' Health Physics ,

Conditions ,

As a result of the original investigation, NSRS identified a need for a program to evaluate unusual health physics conditions with emphasis placed on reduction of exposure potential. The program elements should contain trend analysis of exposures, contamination incidents, incremental increases in dose and dose rates plus a variety of other indicators of problem areas. For this follow-up review, it was detemined that computerized ALARA information system user procedures have been developed. An ALARA engineer was placed at SQN to evaluate plant conditions and identify reasonable methods to reduce radiation exposures. The procedures, reports issued, and trend charts were examined and detemined to satisfy the intent of the recomunendation.

This item is closed (see section IV.I for details).

J. I-82-21-SQN-05, Emphasize safetv-First Pohey to All Employees As a result of the original investigation, NSRS recosmonded that the TVA Board of Directors safety-first policy be impressed upon all employees. For this follow-up review, it was determined that shortly after the thimble tube ejection incident, the TVA Board of Directors issued a memorandum to all TVA employees which expressed the safety-first policy. This item is closed (see section IV.J for details).

K. I-82-21-SQN-06, Practice of Removinz Cap from Vial of Na-24 be Reevaluated As a result of the original investigation, NSRS recommended that the practice of removing the cap from a vial of Wa-24 be reevaluated and the use of a tool be considered. For this follow-up review, NSRS found that the tests requiring the use of NL-24 will not be conducted again at SQN. One procedure had been cancelled and the remaining procedure will be cancelled When open items associated with the test are closed. This item is closed (see section IV.K for details).

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l l L. I-82-21-SQN-07, Adiust Extremity Exposure Records to Reflect '

Reasonable Maximum Exposure .

In the original investigation, NSRS concluded that exposure records for employees involved in the 10 rem extremity exposure incident did l not reflect the probable actual exposure. For this follow-up review, NSRS found that extremity exposure records for the employees involved i

had been revised to show calculated values for reasonable maximum i

exposure. This item is closed (see section IV.L for details).

M. I-82-21-SQN-09. Evaluate Extremity Monitorina Proaram at TVA and

+ particularly 505 j As a result of the original investigation, NSRS recommended that the

' extremity monitoring program be evalusted for the capability of identifying extremity exposure sources and interpreting extremity exposures from all radiation sources encountered with emphasis on 1 seemingly point sources. For this follow-up review, MSRS found that an evaluation of the extremity monitoring program has been performed. The Radiological Protection Plan and plant instructions y

provide for the identification of extremity exposure hazards, ALARA preplanning, and the appropriate use of extremity monitoring devices i{ to ensure that extremity doses are accurately measured. This item is closed (see section IV.M for details).

N. I-84-12-SQN-01, Inadeouste Corrective Measures to Alleviate the i Dearaded Condition of the Thimble Tubes 1

As a result of the original investigation, NSRS recommended that responsibility for overall system operability be assigned to plant

, engineers. This responsibility would be to periodically assess ,

j system performance, operations, maintenance and testing, and to assure problems are promptly identified and corrected. For this follow-up review, NSRS determined that a new procedure was recently

! issued Which identifies the assignment of each plant system to a l designated plant section, e.g., the Reactor Engineering Section is j responsible for the incore flux detectors. Two reactor engineers are currently assigned to this system (one engineer for each reactor unit). The engineers are responsible for performing all the tasks in

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the NSRS recommendation. System status is reviewed on a continuing l

basis. This item is closed (see section IV.E for details).

i' O. I-84-12-SQN-02, Inadequate Survey and Feedback to Field Services Group (FSG) Personnel i As a result of the original investigation, NSRS recosunended that assignments be given to those knowledgable and that they be held j responsible for the success and safety of the operation to be I accomplished. For this follow-up review, it was determined that-i since the thimble tube event, it has been stressed to all staff personnel that they will have responsibility and be held accountable for the success and safety of operations to be accomplished. Based upon the review of documentation and discussions with supervision, I

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NSRS found no reason to doubt that work assignments would not be ' .

given to those most knowledgeable and are available to perform the operation. Planning of activities is comprehensive, as indicated in the unit 2 target schedule with the items receiving close management attention, particularly if they are likely to produce potential problems or delays. This item is closed (see section IV.0 for details).

P. I-84-12-SQN-03, Inadequate Decision Makinx Process As a result of the original investigation, NSRS recommended that management identify and thoroughly evaluate hazards associated with unique activities and that techniques such as systematic hazard analysis methodology be used. For this follow-up review, it was determined that SQN uses the hazard assessment methodology to evaluate the safety of unique operations. NSRS examined several hazard assessment worksheets and they were adequate. Also, it was determined that Management Oversight Risk Tree (MORT) analysis was used to assess the safety of performing the unique activity of entering the unit 1 pressurizer enclosure to do repairs with the unit at full power. The work was subsequently performed safely as the hazard analysis concluded. This item is closed (see section IV.P for details).

Q. I-84-12-SQN-04, Assignment of Work Function to the FSG as an Ordinary Work Activity As a result of the original investigation, NSRS recommended that sufficient time and information be provided to properly plan the activity and that the knowledge and background of workers assigned is adequate. For this follow-up review, it was determined that management has stressed to employees the importance of safety first, advance planning is taking place, maintenance planner positions have been established and staffed, and trained and knowledgeable personnel are being assigned to perform tasks. This item is closed (see section IV.Q for details).

R. I-84-12-SQN-05, Selection of an Inappropriate Instruction for the Control of the Work Activity As a result of the original investigation, NSRS recommended conducting an awareness program to stress the importance of procedure controls, compliance with procedures, the proper chango process for inadequate procedures, and SQN policy as stated in SQA 129. For this follow-up review it was determined that it has been conveyed to plant personnel the importance of compliance with procedures TVA safety-first policy, foremen and craft personnel have been informed on how to use plant instruction change forms, the daily Plant Manager meetings discuss any failure to follow procedures, and the SQA 129 policy has been stressed to employees. This item is closed (see section IV.R for details).

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S'. R-84-12-SQE-06, Inadeouste Job Safety Analysis and Hazards Assessment l

As a result of the original investigation, NSRS reconumended that the l

Job safety analyis program be upgraded; that an effective hazards I assessment methodology be established as a tool to analyze the identified radiological and industrial aspects of the job, the probability of an accident, and the impact on the workers, plant, and the public; and that the recomumendations of the NSRS Report No.

I-82-21-SQE be implemented.

For this follow-up review, it was determined that the remaining open items from report I-82-21-SQE have been satisfactorily resolved (sections IV.H through IV.M of this report); hazards assessment methodology is established and it uses conservative accident assumptions; job safety planning instructions exist and are being used; ALARA preplanning criteria and checklist have been expanded to cover radiological hazards; maintenance planning positions with overall responsibility for job safety have been established and staffed. This item is closed (see section IV.S for details).

T. I-84-12-SQN-07. Inadeouste Field Quality Engineerinz (FOE) Review of ,

Maintenance Request (MR) and Reference Work Instruction '

As a result cf the original investigation, NSRS concluded that the quality of the FQE review process of MRs should be improved to assure l the quality of referenced work instructions, the proper program controls are identified and the instructions,are appropriate for the job. For this follow-up review, it was determined that an evaluation of the MR process was conducted by the Quality-Engineering group.

They identified the need for training of personnel involved in the MR planning process and adjustments were inade to upgrade the QA review program. The training records for QA reviewers, section instruction letters and MR review reports were examined and determined to adequately address the reconumendation. This item is closed (see section IV. T for details).

U. I-84-12-SQN-08, Noncompliance with Recuirements af RWP No. 01-1-00102 As a result of the original investigation, NSRS recommended to emphasize compliance with requirements of RWPs to employees for their own protection. For this follow-up review, it was determined that the RWP cover sheet has been modified to state that entry into containment will be performed in accordance with AI-8 " Access to Containment". AI-8 has been modified to require that the incore detector system be tagged with a hold order prior to issuing an RWP for the lower containment or annulus and specifies that the hold order will be issued to the HP shift supervisor by title. Other modifications to AI-8 were made to eliminate confusion on coordination of maintenance activities and access to the lower containment or annulus. Also, clearance procedure training has been conducted for appropriate personnel as discussed in section IV.V of this report. RWPs and tCI-10. " Minimizing Occupational Radiation Exposures," training has been given to a significant number of SQE personnel. This item is closed (see section IV.U for details).

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V. I-84-12-SQM-09, Noncompliance with Requirements of Section 5.1.4'of

, AI-3. " Clearance Procedures" As a result of the original investigation, NSRS recommended that l

strict compliance with the requirements of AI-3, " Clearance Procedures," be emphasized and enforced. For this follow-up review, it was determined that the following corrective actions were taken to resolve this recommendation: (a) formal training on AI-3 was conducted and only those personnel that passed the exam were included in the revised clearance authorization list in Appendix A to AI-3, (b) AI-8 " Access to Containment" was modified to reauire the shift engineer to issue the hold order clearance on the incore detector system to the HP Supervisor by title and controls access to the lower containment or annulus based upon the status of the incore detector system, (c) hold order concerns have been discussed in outage critique meetings. (d) hcid orders were discussed at crew safety I meetings and (e) planners are instructed to minimize use of hold orders. This item is closed (see section IV.V for details).

W. I-84-12-SQN-10. Modification of Cleaninz Tool Base Supports Without Performina a Technical Evaluation or Testinz i

As a result of the original investigation, WSRS recommended that it be emphasized to the plant staff that changes to tools and equipment affecting work on critical structures, systems and components can be made only after conducting a thorough technical evaluation. For this follow-up review, it was determined that a standard practice SQM 63 for special or modified tooling has been prepared and is being used.

Twenty special tool evaluations have been prepared. This item is closed (see section IV.W for details).

I. I-84-12-SQW-11, Violation of Work Instruction As a result of the original investigation, MSRS recommended that management emphasize that adherence to PORC reviewed, plant manager approved instructions is mandatory and periodic assessment of compliance with instructions should be initated and corrective action taken. For this follow-up review, it was determined that adherence to procedures is being emphasized by SQW management as discussed in sections IV.Q and IV.R of this report; the Plant Manager discusses events infolving failure to adhere to procedures at his daily management meetings; the Quality surveillance Section performs periodic assessments of compliance with instructions. This item is closed (see section IV.I for details).

Y. I-84-12-SQN-12, Lack of Control of Enress capability from Containment As a result of the original investigation, NSRS recommended that a policy and methodology be established'to require an evaluation of the effect on work in progress and notification of affected workers, as necessary, before granting permission to incapacitate egress routes from the reactor containment. In addition it was recommended that-the risks of working in containment and established controls for containment integrity be emphasized to employees. For this follow-up i 10 l

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review, it was determined that the instruction AI-8 " Access to 4

Containment" has been revised to require notification be given to'  !

personnel within containment to use an alternate exit if an airlock 1

! door should be made intentionally inoperable. The actions taken in

response to recommendations I-84-12-SQN-6, - 8, - 9. - 13 , -20 and ,

I-82-21-SQN-5 (sections IV. S, U, V. Z GG, and J of this report)

adequately discuss the employee job safety and awareness aspects of '

{- this recommendation. This item is closed (see section IV. Y for j details).

Z. I-84-12-SQN-13, Breakdown in the ALARA Preplannina Promram

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As a result of the original investigation, NSRS recommended that it l be emphasized to the plant staff that compliance with ALARA j

preplanning requirements as specified in RCI-10. " Minimizing a i Occupational Radiation Exposures," must be accomplished. For this '

} follow-up review, it was' determined that positive corrective action has been taken by: (a) making extensive modifications to RCI-10 to require an ALARA preplan based upon eight criteria rather than the

, previous single criteria and adding an extensive 41 item preplan and post-plan checklist (b) training on RCI-10 and RWP has Deen given to a significant number of SQN employees, and (c) HP participation in PORC subcommittee biennial review of existing and all newly proposed plant instructions and, where appropriate, adding a precaution to have personnel contact HP for applicable RWP, AI-33 shielding and ALARA peoplanning. This item is closed (see section IV.Z for i details). '

t AA. I-84-12-SQN-14 Need for Formal Documentation for Upper-plant Manamoment Approval to Work in Radiation Dose Rate Fields Greater j Than 50 Rem / Hour

, As a result of the original investigation, NSRS recommended that SQN

establish formal requirements and provide a method to document the

authorization to work in dose rate fields greater than 50 rem / hour, j For this follow-up review, it was determined that Radiological Control Instruction RCI-14 " Radiation Work Permit (RWP) Program", has been revised to require formal documentation of the review of all RWPs when the work area dose rate equals or exceeds 50 REM / hour or prior to any entry inside the polar crane wall when the reactor is at power. This item is closed (see section IV.AA for details)

BB. I-84-12-SQN-15, Availability of Communications Followinn the Accident As a result of the original investigation, NSRS recosmonded that anytime the telephone is out of service in the airlock, alternate

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communication methods should be considered. Also, the availability  ;

of communications should be considered durin, the performance of job l safety analysis and job planning. For this_ follow-up review, it was l determined that the instruction AI-8, " Access to Containment," was l modified to require: (a) the Public Safety Officer unlocking the airlock ensure that the phone inside the airlock is checked.for 11 l I

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s proper operation prior to the first person entering containment and ,

i fill out a data sheet related to phone operability, (b) if the phone is not working an immediate attention maintenance request is.

initiated to Electrical Maintenance for repair, and (c) entry into containment during the period the phone is out of service shall be approved by individual supervision or the Shift Engineer when the

i. supervisor is not present.

) Also, the RCI-10. " Minimizing Occupational Radiation Exposure," was i revised to add a projob ALARA planning report checklist. This i checklist requires that a determination be made on whether special l

communications equipment is needed to enable workers to connunicate effectively. This item is closed (see section IV.BB for details),

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, CC. I-84-12-SQN-16. Effective Cleaninz of the Thimble Tubes by WUS Corporation As a result of the original investigation, NSRS reconamnded that SQN advise WBN of the effectiveness of the NUS cleaning method over the Teleflex method. For this follow-up review, it was determined that the SQN Plant Manager informed WBN of the NUS cleaning method. The WBN instruction MI-94.3, "Incore Flux Thimble Cleaning and Lubrication," has been modified to require the,use of the NUS i

cleaning equipment and methods. This item is closed (see section IV.CC for details),

i DD. I-84-12-SQN-17, Poor Quality Cleaninz Procedures and Inadequate PORC Review As a result of the original investigation, NSES had expressed

, concerns with the adequacy of maintenance instructions and the PORC procedure review process. It was recommended that an evaluation be

, made of the PORC procedure review process with consideration given to supplementing the review process with expert subcommittees; cancel SMI-0-94-I; do not use EMI-0-94-2 until it has been revised to i include quality elements; perform a generic review of all maintenance and special maintenance instructions to ensure adequacy. For this j follow-up review, it was determined that.SMI-0-94-1 and SMI-0-94-2 l were cancelled and a thimble tube cleaning procedure MI-1.10 "Incore Flux Thimble Cleaning and Lubrication," issued. The Plant Manager stated that an evaluation of the PORC procedure process has been conducted and that the work load needed to be reduced and that steps

have been taken to reduce this load. Subcommittees are being~used

! for the procedure review process. This item is closed (see section i IV.DD for details).

i EE. I-84-12-SQN-18, Noncompliance with Serious Accident Reportina and Accident Scene Preservation Requirements As a result of the original investigation, NSRS recommanded that

! corrective action be taken to ensure future compliance with TVA l established requirements for accident reporting and scene

preservation. For this follow-up review, it was determined that an I 12 l

.._ ~. _ . . _ _ _ __ __ _ _ _ _ _ _ . __ ._ _ _ _ _ _ _ _ _ _ . _

1 ONP procedure for serious incident investigations was issued in May 1985 and a SQW site procedure SQS 29 issued in ' July 1985. These documents satisfactorily address accident reporting and scene

preservation requirements. This item is closed (see section IV.EE for details).

FF. I-84-12-SQN-19. Limited NUC PR Accident Investination

! As a result of the original investigation, NSRS recommended'that in

future accident investigations potential conflict of interest should

be avoided; the investigation initiated as soon as possible eith r sufficient time for its conduct; it should encompass all aspects of the accident; the ONp recommendation that consideration be given to ~'

l leaving the inner door open should be deleted. For this follow-up i review,'it was determined that the ONP procedure for serious incident investigations and the SQN site procedure SQS 29 adequately address l these items.

The recommendation for leaving the inner door open was not j implemented since AI-8, " Access to Containment," was appropriately

modified. This item is closed (see section IV.FF for details).

GG. I-84-12-SQN-20, Needed Reemphasis on the TVA and SON Eumlovee

Expression of Concerns for Safety and Safetv-First policies

] ~As a result of the original investigation, it was determined that during the thimble tube cleaning the employees did not relate their i increasing concerns for the safety of the job to upper management.

! It was recommended that it should be emphasized to all SQN employees l that they are responsible for voicing their views on safety and that all supervisors, engineers, and foremen must evaluate responsible concerns expressed to them. For this follow-up review it was determined that the recently implemented employee concern program j satisfactorily resolves this reconumendation. This item is closed (see section IV.GG for details),

i HH. I-84-12-SQN-21, Ineffective SON ISEG Activities As a result of the original investigation, NSRS recommended that SQN reorganize or reassign functions as necessary to provide ISEG personnel adequate independence from line responsibility and pressure and to limit their functions to ISEG type duties as required by the Technical Specifications. For this follow-up review, it was '

deterinined that the ISEG/ Compliance Staff has six engineers and one supervisor that perform the ISEG/ Compliance function. Many of the compliance functions, e.g., LER preparation, potentially reportable events review, scram investigations, etc., are ISEG-type functions.

Based upon the review of several of the ISEG reports and discussions with several engineers, nothing suggested that the ISgG-type work was being compromised by the dual responsibilities. Also, proposed 3_ Technical Specification changes with justification have been submitted to the NRC that'show the ISEG/ Compliance Staff reporting to the Site. Director with the staff having dual responsibilities, per

! discussions with plant management, this current reporting arrangement i j has been discussed with NRC personnel. In addition, the ONp '

organization changes are being 13

6

, considered that could impact the ISEC reporting arrangements. .The .

resolution of the proposed Technical Specification change with NRC

! combined with potential revisions to the ONP organization will result I in resolution of this item. Because positive action has been taken and a resolution with WRC is being pursued at the highest levels of

l. TVA nuclear power management, this item is closed (see section IV.HH I

for details).

f

! II. I-84-12-SQN-22. Sinnificant Breakdown in the SON Procedure Process for Maintenance Activities As a result of the original investigation, NSRS recommended that the l procedural process for maintonance activities be throughly l evaluated. Corrective actions should be initiated to: (1) improve i the knowledge of personnel preparing and using procedures as to what l constitutes an appropriate procedure, quality elements to be incorporated into a procedure, and the change process for the l procedures; (2) improve quality of-PORC and biennial reviews, and l (3) compliance with procedures. For this follow-up review, l discussions were held with supervisors and engineers in Maintenance and Quality Assurance and documents related to the preparation, review, and implementation of maintenance' instructions were examined. It was determined that the following actions have been taken to improve the procedural process for maintenance activities:

(1) meetings were held with craft and foreman to inform them how to use plant instruction change forms, (2) reviews of procedures are being made in the draft stage. (3) the craft are now required to review draft instructions or new revisions, (4) a draf t procedure writing guide has been developed by Mechanical Maintenance, (5) a procedures review checklist has been developed and is being used, and (6) a comunitment has been made to WRC to review all Maintenance Instructions with a fully developed checklist by July 1987.

Additional discussion on PORC reviews is provided in section IV.DD of this report. Emphasis c.n adherence to procedures is addressed in sections IV.Q and IV.R of this report. This item is closed (see section IV.II for details).

JJ. I-84-12-SQM-23 Inadequate Reportinst of the Event to NRC As a result of the original investigation NSRS recommended that SQW revise the licensing event report (LER) to reflect the true nature of the leak, the adequacy and violation of SMI-0-94-1, and the effective long term corrective action. For this follow-up review, it was determined that the ESRS thimble tube investigation report l

I-84-12-SQW and the IUC PR response were attached to a revised LER

[

submitted to NRC thus making these documents part of the LER. The combination of these documents address the nature of the leak, l procedure inadequacies and proposed corrective action. This item is closed (see section IV.JJ for details).

KK. R-84-17-NPS-02, Lack of Approval of Onsite Vendor Services at SON In the original review, NSRS recommended that SQN develop and implement a program that satisfies the requirement and intent of 0QAM, part III, section 2.1, paragraph 10. The original review cited three examples of vendor services for which no QA documentation was 14

provided to demonstrate that the work was accomplished in accordance with QA requirements. Since no documentation was initially provided, NSRS assumed none existed. For this follow-up review, subsequent documentation of the three cited cases was provided to NSRS. Review of this documentation and telephone conversations with one of the vendors was sufficient to demonstrate that proper QA control was applied to the vendors and adequately monitored by SQN. This item is closed (see section IV. KK for details).

LL. R-85-02-SQN/WBN-01 (NUC PR) Office-Wide Awareness Bulletin for Tube Fittina Maintenance Activities t

In the original review, NSRS recommended that a ONP office-wide awareness bulletin be sent to the nuclear plants which discusses:

tube fitting design; assembly, reassembly, and inspection criteria; Policy on interchanging components; failure modes; hazards involved in working on pressurized fittings; precautionary measures and that the bulletin be incorporated into a permanent instruction at each plant for new employees. For this follow-up review, it was determined that a safety awareness bulletin was sent to WBN, ELN, SQN and BFN. The bulletin addresses many of the recommendations and identified the prime elements contained in a tube fitting installation training program that had been prepared by the POTC.

i All of the plants are committed to provide this training to the crafts persons working with compressed fittings by virtue of the TVA commitment to obtain INPO accreditation of the craft training program. The active nuclear sites are in various stages of completion in providing this training. The combination of the awareness bulletin, the POTC training programs, and INPO accreditation program satifies all of the elements of this recommendation. This item is closed (see section IV.LL for details).

MM. R-85-02-SQN/WBN-02, Maintenance. Operatina, and Test Instructions In the original review, NSRS concluded that SQN instructions were not sufficiently clear and did not include sufficient 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 i

i increased while the thimble tubes are disconnected from the overhead path transfer system. The latter recommendation was intended to preclude the ejection of a thimble tube in the event of failure of a high pressure seal. For this follow-up review, NSRS determined that several recommendations had been incorporated and others were being i addressed in proposed precedure revicions. This item remains open I pending completion of the following actions (see section IV. MM for details).

l 1. Issuance of the proposed MI-1.11. " Thimble Tube Installation,"

! which will replace SMI-1-94-5 and addresses several of the

original recommendations.

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Issuance of the proposed revision to SMI-0-94-3 that requ}re's 2.

the use of an appropriate thread lubricant, and cautions againsf.

allowing fitting bodies to turn.

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

j 4. Revision of appropriate instructions to preclude pressurizing j the primary system with the thimble tubes disconnected from the overhead path transfer system or at least preclude any work on

! the seals with the primary system pressurized above atmospheric and the thimble tubes disconnected from the overhead path

transfer system.

i l NN. R-85-03-NPS-01, Inadeouste Definition of Responsibility l

l In the original review, ESRS concluded that the responsibility for determining the identification and avaC2bility of spare parts was i not clearly defined in procedures. SQN reaponded that this l responsibiltiy was in the job descriptions for maintenance planners.

l For this follow-up review, NSRS determined that the maintenance i planner job descriptions do include this responsibility and that the

! maintenance planners were aware of their responsibilities. This item i is closed (see section IV.WN for details).

! 00. R-85-03-NPS-04, ASME.Section II Postmaintenance Valve Testina - SON t

l In the original review, NSRS determined that the Instrument Maintenance Section did not identify the need for ASME Section II valve testing and recosumended training in the Section XI pump and valve program. For this follow-up review, NSRS determined that an l appropriate annual training course had been implemented and that Instrument Maintenance Section planners were aware of ASME section XI *

requirements for components within the responsibility of the l

Instrument Maintenance Section. This item is closed (see section IV.00 for denils).

PP. R-85-03-NPS-06, Postmaintenance Testina Pronram-Generic In the original review, NSRS determined that no guidelines were available to ensure that postmaintenance tests verified that the component or system still functioned as designed. For this follow-up review, NSRS determined that Standard Practice SQM-2, " Maintenance Management System," has been revised to include appropriate criteria l for postmaintenance testing, planners were aware of the requirements,

! and postmaintenance tests were being specified in maintenance

, requests. This item is closed for SQN (see section IV.PP for  !

! details).

i f

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I 1

. 1 QQ. R-85-03-NPS-07, common Mode Failure-Generic In the original review, MSRS determined that the Mechanical Maintenance Section had no program to address common mode failure.

For this follow-up review. NSRS determined that Mechanical Maintenance Section Instruction Letter MMSL-A36, " Common Mode Failure - Maintenance Initiated," has been issued. MMSL-A36 addresses the requirements of the NQAM with two exceptions. This item remains open for SQN pending revision of MMSL-A36 to address the role of calibrated tools in potential common mode failures and to meet the intent of " redundancy of people" as stated in the NQAM (see section IV.QQ for details).

RR. R-85-03-NPS-08, Surveillance of Maintenance Program-Generic In the original review, NSRS determined that surveillance of maintenance activities was inadequate and recommended more indepth surveillance including reviews for proper CSSC classification, postmaintenance testing, ASME Section XI testing, and common mode failure. For this follow-up review, NSRS determined that all the items were addressed in Management Review Guidelines except common mode failure, and that no surveillance checklist had been prepared for postmaintenance testing. This item remains open pending revision of the surveillance program to include common mode failure, issuance of the postmaintenance testing surveillance checklist, and NSRS review of the implementation of surveillances on the maintenance program and postmaintenance testing (see section IV.RR for details).

IV. DETAILS A. R-80-03-NUC PR-C1, Additional Training for STA In the original review (reference A.1), NSRS recommended that additional training be provided to STAS to assure cognizance of existing requirements for reviewing the shift engineer journal. For this follow-up review, it was determined that the on-the-job training being provided to the STAS adequately defines their responsibilities for reviewing logs and journals prior to and af ter assuming shift duty.

The SNP Engineering Instruction Letter ES SIL A11. " Station Shift Technical Advisor Training," establishes the requirements for the on-the-job training for the routine duties, duties prior to assuming shift, delineates the responsibility for administering this training, ,

and defines record retention. The purpose of the training is to  !

provide the STA with the skills necessary to correctly perform these

, duties without assistance when filling the on-shift STA positio-1 l

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Upon completion of the training, the trainee will be able to . .

successfully (not a total list): (1) state which Adminstrative Instruction defines the responsibilities of the STA, (2) list the routine duties of the STA, and (3) list the requirements for shift turnover and state what actions should be taken by both parties to ensure essential information is transferred. The training documents related to these administrative responsibilities are AI-27. " Shift Technical Advisor" and AI-5, " Shift and Relief Turnover",

respectively. The portions of these instructions related to journal review follows:

Section III. 7 of AI-27 states:

Review of reports, technical information, and other related nuclear experience review material in accordance with SQA-26; review of STA's and shift {

engineers daily journals and night order book in accordance with AI-5; and regular review of appropriate control room logs and/or daily journals in order to maintain cognizance of each unit.

Section 2.2.4 of AI-5 states:

Operators (all classifications) and STA's--Transfer of Authority and Responsibility: Oncoming operating personnel shall be responsible to acquaint themselves with the equipment status and any activities under their jurisdiction before assuming the duty for the shift. As a minimum it shall include reviewing the journal entries back to his last shift worked pr back five (5) calender days (7 days for STA's), whichever is less; observance of control boards, alarm panels, etc.; determination of plant status as related to technical specifications as. . . .

Thus, training and adherence to these instructions ensures review of the shift engineers journal. Based upon discussions with an STA and Reactor Engineering Supervisor, review and signoff of the shif t engineer journal is not required prior to assuming shift responsibility. However, once responsibility is assumed, the plant status and review of logs and journais (including shif t engineer) is performed. Based upon the review of the STA on-the-job training instruction, and discussions with the STA and the.

Reactor Engineering Supervisor it is concluded that adequate training is provided to assure cognizance of and compliance with the the requirement to review the Shift Engineer journal. This item is closed.

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B. R-80-03-NUC PR-C2, Adecuate STA Trainint Records In the original review (reference A.1), NSRS reconumended 1 that a review of STA training records be made to assure l that adequate records are available to demonstrate that training has been completed. For this follow-up review, it was determined that the shift technical advisor training documentation requirements as specified in section 4.3 of i ES SIL All, " Station Shift Technical Advisor Training,"

requires:

. When the STA trainee has demonstrated proficiency in l meeting the objectives by performing the various duties without supervision, the person administering

the training will document the trainee's satisfactory i performance by completing a signoff sheet similar to '

l Attachment 2 (page 10).- The Power Operations Training i Center (POTC) will retain the completed signoff sheet t

! in their flies for verification that the trainee has' i satisfactorily completed training and is certified by

! a qualified STA to be capable of assuming the STA l shift.

i Several completed signoff sheets obtained from POTC were reviewed. ,

! They verify that individuals received on-the-job STA training and .

t

! certified that they demonstrated satisfactory proficiency to be j qualified to assume the STA shift. Conversations with the Reactor i Engineering Supervisor and an STA confirm the training was provided j and that the STA possessed a good working knowledge of the review requirements provided in the training. This item is closed.

]

C R-80-05-SQE-04, Electrical Deficiencies i

Part 4 2

i In the original review (reference A.2), NSES found four specific 1 electrical deficiencies during a walkdown inspection of the plant.

! Three of these items were corrected and verified complete in NSRS ,

follow-up review R-80-11-SQN. The last item remained open. This

! item concerned exposed cables between penetration 25 and the cable j tray leading to the penetration in the annulus. For this follow-up review NSRS inspected the cables and cable tray at penetration 25 I (unit 2) and found the cable tray cover in place and the cables l'

properly coated with mastic between the cable tray and'the penetration. Part A of this item is closed.

i 4

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part B ' -

In the original review, NSRS concluded from the electrical deficiencies found that configuration control was inadequate, and l made the following reconumendation:

i Revise the existing configuration control i

program including appropriate instructions to j require more frequent and indepth inspections i j such that the entire plant has been inspected once per refueling cycle.

l For this follow-up review, NSRS examined the configuration control

! programs in place at the plant and reviewed the status of the TVA l changeover to the Configuration Control Drawing system at.SQN. A number of actions have been taken which improve the confidence that

the "as-constructed" drawings and the actual plant configuration of 1 CSSC systems agree. Some of these are

1. Control of temporary alterations has been tightened over the

years and a periodic PORC review of outstanding temporary i alterations is required (AI-9, " Control of Temporary Alterations

! and Use of the Temporary Alterations Order").

2. Control of plant modifications (Al-19, part IV " plant Nodifications: After Licensing") requires marking the required control drawings-as soon as the modification is field complete

and provides for marking the drawings for a partially completed modification if the system is to be operated.

i

! 3. All personnel are charged with immediately reporting j discrepancies they find between the plant configuration and

! "as-constructed" drawings (AI-25, part I, " Drawing Control After j Unit Licensing").

l 4. NI-6.20, " Configuration Control During Maintenance Activities,"

i provides a simple method of controlling temporary conditions

during maintenance activities as an alternative to the Temporary j Alteration Control Forms (TACFs). This reduces the number of TACFs, which makes that program easier to track and makes j control of temporary conditions during maintenance easier and j therefore less likely to be subverted.

}

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

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

l 7. Drawings with areas marked " incomplete" were updated with j information obtained by researching construction work packages.

l 20 i

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'l The preceding actions are examples of things that show that '

configuration control is notably improved since the 1980 review that l resulted in this recomunendation. The TVA wide change to the i Configuration Control Drawing system currently being pursued should l improve the situation further. Phase I of this effort is to

establish the configuration baseline (which included 3CN plus

modification work) for the primary drawings (selected drawings on

critical systems) on the "as-constructed" drawings without the aid of l

system walkdowns. At the time of this review, a list of drawings considered necessary for CSSC configuration control had been i

I identified by the plant. The intent is to complete Phase I of the new program for these drawings prior to restart. Completion of this

effort will also increase confidence in the configuration control program. In establishing the configuration baseline, NSRS considers it important to verify that the plant configuration and the

"as-constructed" drawings agree by performing walkdown inspections.

j Adequate confidence in the CSSC configuration baseline may be i achieved by walkdowns of selected systems, if so, walkdowns of all l CSSC systems would not be necessary.

)i In consideration of the considerable improvement in configuration i

control since 1980, NSRS no longer considers it necessary to verify

! plant configuration every refueling outage. Part B of this item j rema".na open pending verification of the configuration baseline. To achieve this, NSRS recommends the following actions be taken prior to j restart.

i 1

{ 1. Completion of Phase I of the plan for conversion to i configuration Control Drawings for those drawings previously l identified by the plant as necesssey for CSSC configuration j control.

i l 2. Selection by the plant, of a number of CSSC systems for complete l walkdown to verify that the actual physical configuration agrees +

{ with-the CSSC configuration control drawings verified as part of

the Phase I effort.

l 3. Walkdown of the selected systems and correction of any i discrepancies found. If significant problems are found,

{ additions 1 systems should.be selected for walkdown.

j Part c 1

1 In the original review, NSRS recommended that programs be implemented

' or revised for all employees to emphasize the need for and maintenance of configuration control. For this follow-up review,

! NSRS reviewed various instructions, training programs, surveillance j reports, and maintenance requests, and interviewed personnel. NSRS '

! found that employees are oriented and/or trained to the procedures ,

j and instructions that affect them, such as MI-6.20; AI-25, part 1; and AI-37, " Independent Verification." Problems with the i implementation of various aspects of configuration control were j identified in surveillances, but none of them were attributable to i

lack of awareness of the need for and maintenance of configuration l control. NSRS concludes that employee awareness of configuration

{' control is adequate. Part C of this item is closed.

i 21 i

1 e e D. R-81-07-SQN-07, Unreviewed Temporary Alteration Control Forms . .

Part 1 The NSRS found that AI-9, " Control of Temporary Alterations and Use of the Temporary Alteration Order," did not agree with DPM N07311 in two important areas and recommended that AI-9 be revised. AI-9 did not require PORC review and Plant Superintendent approval of

temporary alterations, nor did it require issuance of a design change request for CSSC temporary alterations that are to remain in effect over 60 days.

Since the original review, DPM N07311 has been replaced by NQAM, part II, section 6.4. AI-9, revision 19, reflects the requirements noted above as stated in NQAM, part II, section 6.4, dated November 5,1984, although the requirements have changed since the original review, and AI-9 is not identical to the NQAM. The NQAM states; i

A DCR or FCR shall be submitted if an alteration is to remain in effect for more than 60 days unless the TA 6

is to be removed prior to operation of the affected system (s).

AI-9 states:

l A DCR or FCR shall be submitted if an alteration is to i remain in effect for more than 60 days unless the TA is to be removed prior to operation of the affected system (s) or the TA is to be removed under an existing ECN.

, The additional exception in AI-9 does not change the intent. Part 1

! of this item is closed.

f Part 2 The NSRS found that many unit 2 CSSC temporary alterations had been in effect for nearly a year at the time of the review and recommended that the status of outstanding unit 2 CSSC temporary alterations be reviewed prior to fuel loading.

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• During this follow'-up review, NSRS found no evidence that the status

! of outstanding temporary alterations had been reviewed prior to fuel

{ loading. However, because Unit 2 fuel Loading has occurred, review l emphasis was placed on actions that have been and are being pursued j since fuel loading. The status of.all outstanding temporary' j' alterations is reviewed by PORC every six months as required by the

! NQAM and AI-9. In addition, a monthly TACF Status Report is issued j to section supervisors and others. This report includes a graph of j the number of outstanding temporary alterations versus time with

goals identified, and serves to keep managers and supervisors aware of their temporary alterations. SQN also committed to INPO to close all temporary alterations established prior.to 1984 (148 outstanding at the time) by the unit 1 cycle 4 refueling outage (next refueling). The November 1985 and January 1986 Monthly TACF Status Reports show 96 and 84 temporary alterations in this category .

i respectively, showing that progress is being made on old temporary j alterations as well as all temporary alterations. Because of the apparent emphasis on closing temporary alterations and the steady 3

downward trends in all temporary alteratiors, CSSC temporary alterations, and old temporary alterations, Part 2 of this item is

closed, l'

{ E. R-82-04-NPS-01, Containment Sorav Test Line at SON and WBN In the original review (reference A.6), NSRS recosmonded that automatic isolation of the test line (recirculation line to refueling i

water storage tank) at SQN and WBN be provided to isolate this line j if an accident occurs whenever the containment spray system is 1 required to be operable. NSRS previously reviewed revision 16 of SQW i

SI-37 "Contain:nent Spray Pump Test," which had the added requirement to station an AUO near the containment spray system valves in question and to be in communication by telephone or other means with j the control room prior to and during the test. This action satisfied j the intent of the recommendation. The WBN portion was not satisfied,

thus the item remained open. For this follow-up review, it was

! determined that WBN has modified instruction SI-4.0.5.72-P, I

" Containment Spray Pump Test," to require an AOU to be stationed near j the recirculation line valves in question and to be in communication by telephone or other appropriate means with the main control room 3} prior to and during the performance of the test. This change

! satisfies the intent of the recommendation. This item is closed, j

j F. 1-82-20-SQN-01 Administration of KI to Plant Personnel 4

l In the original review (reference A.7), NSRS found that SQN-IP-20 i reconnended that field monitoring, team members take KI when their j pocket dosimeter reads 25 arem, but MSEC-IP-9 made no KI recmunendations. NSRS recommended that consistent guidance be i provided for the administration of KI. For this follow-up review, .*

j NSRS compared SQW-IP-20, revision 3, and CECC-IP-9 (replaced '

j MSEC-IP-9), revision 4, and found that they provide the same guidance j for administration of KI. This item is closed.

i i ,

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G. I-82-20-SQN-02, Usarade of Field Team Van . ,

In the original review (reference A.7), NSRS recommended that ONP and the Radiological Health Staff (RHS) upgrade the field team vans to include a permanently mounted seat in the rear and compartmentalized, labeled equipment storage. RHS responded that all the vans would be renovated by July 1983. For this follow-up review, NSRS inspected-the SQN van and found it had been upgraded as reconmended. Some of the stick-on labels were beginning to peel off the shelves. The Health Physics Supervisor initiated correction of this problum by submitting Work Request B101960 to replace the stick-on labels with painted-on labels. HPSIL-28 " Quarterly Emergency Van Inventory,"

requires an inventory and restocking, if necessary, once per quarter and each time the van is used. NSRS compared the van contents to the most recent inventory and found only minor discopancies which could have been due to the fact that a plant "open house" was held the previous day and many people had toured the van. These minor discrepancies were corrected by Health Physics personnel. This item is closed.

H. I-82-21-SQN-02 Emphasize pre-Job planning In the original review (reference A.11), NSRS recommended that additional emphasis be placed upon pre-job planning and procedure development and review by both the Nuclear Central Office (NCO) and SQN to ensure that hazards associated with a job are identified and reduced to a level of risk acceptable to management and that proper equipment is assembled and in working order prior to the start of a job. The NUC PR tasponse dated (reference A.15) stated that, as reconmended, they would place additional emphasis on pre-job planning, procedural development, and review by both the plant and the NCO, as appropriate, to ensure that the safety hazards associated with the job are reduced to an acceptable level of risk. The response also stated that this is a continuing action and no follow-up would be provided.

For this follow-up review, it was determined that since then new positions of maintenance planners have been created at SQN. Four job descriptions were reviewed. One was approved May 5, 1985 and the others April 15, 1985. All contained the following duties and responsibilities in addition'to other work details:

The incumbent is responsible for developing detailed work plans for individual maintenance activities at a nuclear power plant, including detailing the job sequences required to perform the tasks. The maintenance activities include preventive and corrective maintenance, forced and scheduled outage activities, and other work such as inspection of plant equipment. In the development of the respective maintenace work plan, the incumbent ensures that the required drawings, technical manuals, work instructions, parts / materials, special tools, and 24

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i * ,

' required preconditions are identified and that support craft and services required for the accomplistunent of the maintenance work are coordinated before the job is started. Inspects the worksite prior to planning the j

maintenance work as necessary to verify problem is '

correctly identified and to determine the need for special tools, permits, unusual conditions,-

{. preparatory work requirements, etc.

The incumbent is responsible for protecting the health j and safety of employees and for safeguarding TVA i

property. This is accomplished by including proper safety precautione and instructions in the peoplanned

, work instructions. Responsibilities include the permanent removal of employees or other persons from situations Where hasards are not adequately i controlled; understanding and implementing the provisions of the "TVA Occupational Health and Safety program plan," TVA code VIII OCCUPATIONAL HEALTH AND SAFETY, and Office and division administrative safety procedures; assuring employee compliance with

) established safety practices and procedures; and planning, supporting, and promoting health and safety

{ as an integral element of the TVA mission.

1 j

Hazards associated with a job being identified and reduced to a level of risk acceptable to management is addressed in the details of recossmendation I-82-21-SQN-04, section IV.I of this report. In addition there is a hasard assessment worksheet in HCI-029 "Workplace Hazard Assessment " which can be used to establish priorities to correct identified hasseds. The ALARA peoplanning aspects of jobs has also been significantly improved. This is discussed in section IV.Z (I-84-12-8QN-13) of this report. This item.

l is closed.

j -

I.

I-82-21-SQN-04, Establish program to Evaluate Unusual Health phystes l ,

Conditions '

t In the original investigation (reference A.11), NSRS recomunended that

~

a program be established to evaluate unusual health physics i

conditions or results with emphasis being placed upon the reduction of exposure potential. program elements should contain trend analysis of exposures, contamination incidents, incremental increases j in dose, dose rates, contamination, and a variety of other indicators

of problem areas.

I I The NUC PR response (reference A.23) stated that the situation was principally caused by the unavailability of exposure information and f the responsibility for high dose exposure review not being assigned l to a specific individual. A computerised ALARA infonnation system l and user procedures were being developed, and an ALARA engineer was j

being placed at each plant to evaluate plant conditions and to identify reasonable methods to reduce radiation exposures.

l

f. 25

=

l . ,

For this follow-up review, procedures were reviewed and discussions ,

were held with the ALARA engineer and supervisors in the Health Physics area. It was confirmed that an ALARA program had been established and implemented. The Radiation Protection Plan, revision 2 (reference B.41), section 1.2.1.c. provides for the l establishment of a program to encourage workers to suggest i

improvements in the radiation protections program. The " Radiological l Hygiene Program " RCI-1, does not include the ALARA program but refers to the ALARA suggestion program described in SQA-145, "As Low l As Reasonably Achievable (ALARA) Suggestion Program."Section V of l RCI-1 notes that all major radiological incident reports (RIRs) were l to be reported to the NSRS. The ALARA suggestions program is in

effect but no suggestions had been received during 1986 at the time

! of this follow-up review. The SQN Objectives in Plant Operations for fiscal year 1986 includes goals for reduction in the personnel contaminations reports, RIRs, and radiation exposure.

The use of ALARA in planning is contained in RCI-10. " Minimizing Occupational Radiation Exposure." Seccion III states in part:

Included in the program is a computerized ALARA dose tracking system developed for each plant to actively track radiation exposure. This system allows for retrieval of dose icformation to detect variances between estimated and actual dose and to perform trend l analysis so that corrective action may be taken to minimize radiation exposure.

Included in the RCI is a pre and postjob ALARA plan and checklist.

Details of the contents of ALARA reports is given in Health Physics Section Instruction Letter HPSIL-25. "ALARA Program." Reports are issued daily, monthly, annually, and post-outage. The Health Physics Section monthly report January 1986 was reviewed. The report listed the following:

I o number of RWPs issued o radioactive contamination in clean areas o respirator training o whole body counting 0 number of internal contaminations o TLD Badging o personnel radiation exposure o decontamination activities o ALARA activities Also reviewed was the annual trend analysis of estimated section exposure dated February 6,1986. The trend charts for the year 1985 were examined. With some exceptions these were monthly plots of the following:

26 l

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

i t

i .

• o RWP discrepancy Reports ,

o personnel contamination reports

) o radiological incident reports

o average dose per worker from RWPs o average dose per worker from TLDs i o quarterly RWP vs. TLD man-res *

{: o cumulative RWP vs. TLD man-rem 4

o average dose per worker from RWPs o average RWP man-hours per worker o average noble gas skin dose per worker o total man-ren '

{i o total noble' gas skin dose

.o total RWP man-hours 4

o total RWP entries o average millirem per RWP entry i

' o average RWP man-hours per RWP entry o average millirem per RWP man-hour o average noble gas skin dose per RWP entry It is concluded that the ALARA program contains the 3

elmaants of the recommendation to evaluate unusual health 1 physics conditions with emphasis on reduction of exposure potential. The program contains trend analysis, contamination incidents, incremental increases in dose. '

dose ratas, contamination and a variety of other indicators of problem areas. The instructions exist-to define the tasks and data is being taken and evaluated. This item is closed.

, J I-82-21-SQW-05, Emmhasize Safetv-First Policy to All

! Reployees

( In the original investigation (reference A.11), NSRS l

recommended that the safety-first policy of the'TVA Board

' of Directors be impressed upon all employees. It was i

recommended that all Health Physics personnel at SQW, all sponsoring groups at SQN, and all Nuclear Central Office groups perfocining work at SQE should receive specific, forinalised instruction on the meaning of safety-first. {

Nuclear Power responded (reference A.15) that it had in place a program which emphasized that safety had priority i

over production. For this follow-up review, NSRS noted ,

that there are in place a number of programs and methods

! where safety is emphasized. Following the incore i

instnanentation thimble tube ejection accident on April 19,

1984, the TVA Board of Directors issued a memorandum to all i

TVA employees on Apell 30; 1984. The subject was "TVA Policy on Reporting Nuclear Safety Matters" in which the ,

safety-first policy was expressed. The Administrative

,' Instruction Power and Engineering "gxpression of Employee '

Views" was revised and issued June 11, 1985, i

1 i 27 i 1 .

i f l  ?

Section V.A.1 deals with Occupationational Health and Safety Ispues ,

and V.A.2 with Nuclear Safety Issues.

A directive for implementation of the new TVA Employee Concern Program (ECP) by February 1, 1986, was issued by the Manager of Nuclear Power. This ECF emphasizes safety in all phases of the employees work. The ECF site representative for SQN comunenced duties on February 3, 1986. Personnel at SQW attended orientation training sessions before February 14, 1986. Posters have been placed in various locations and all new employees receive information on the TVA ECP. Additional programs emphasizing safety and the safety first policy are discussed in section IV.GG (I-84-12-SQN-20) of this report. This item is closed.

K. I-82-21-SQN-06, practice of Removinz Cao from Vial of Na-24 be Reevaluated In the original investigation (reference A.11), NSRS found that a 10 rem extremity exposure occurred in part because a vial of Na-24 was opened by hand. NgRS recomunended that the practice of removing the cap be reevaluated and serious consideration be given to using a tool for cap removal.

The Nuclear Power response was:

This item has been incorporated into the sequoyah turbogenerator acceptance test procedures along with other measures to decrease personnel exposure. The same central office group is responsible for performing this test at all units regardless of plant location so that the lessons learned will extend beyond Sequoyah. This revised acceptance test is being distributed to Watts Bar for their consideration should a comparable test be performed at that facility. This action is complete and no further reports will be provided.

For this follow-up review, NSRS found that the actions indicated in the response were not cospleted. One of the test procedures in question, SQ-STEAR-INST 82-12. " Turbine Benchmark Radioactive Tracer Test Unit 1," was cancelled rather than revised. The other procedure, SU-10.2, " Steam Generator Moisture Carryover Measurement.

Units 1 & 2," was not revised or cancelled. The Reactor Engineering Supervisor stated that SQN elected not to perform the acceptance tests on unit 2, and therefore SU-10.2 was not, and will not, be used again. He also stated that when some open items were closed, he intends to cancel SU-10.2. This item is closed.

28

. _ _ _ _ _ _ _ . .___. _ _ __. _ _ _ _ _ _ _ _ _ __ __. _ . ~. ___

I a .

I i NSRS checked the WBN acceptance test program and found that the

{ corresponding procedure did not contain specific instructions about '

opening the vial. h WBN engineering personnel were aware of the 4

problem and stated that the procedure had not been revised because the Na-24 supplier was revising the packaging and shipping container ,

! and had not specified the final design. The WBN procedure is to be '

} revised as necessary depending on the design of the packaging / shipping 4

container.

l L. I-82-21-SQN-07 Adjust Extremity Exposure Records to Reflect j Reasonable Maximum Exposure i  :

1 In the original investigation (reference A.11), NSRS found that the l exposure records for employees involved in the 10 rem extremity

! exposure incident did not reflect the probable actual exposure. NSRS recommended that the extremity exposure records for Employees A, C, E, and H be adjusted to reflect the reasonable maximum exposure i calculated by NSRS, and the reasonable maximum exposure be calculated i

for esployees B D, and F and their records adjusted accordingly.

l For this follow-up review, NSRS examined'the " Current Occupational External Radiation Exposure" printout dated February 19, 1986, for the extremity dose recorded for the esployees in question, h records for employees C, E, and H had been changed as recommended.

An error was found in the calculation for employee A in the report, the proper value calculated, and the record revised accordingly. h records for employees B, D, and F were revised to reflect a i calculated reasonable maximum exposure, as requested. This item is closed.

i l M. I-82-21-SQN-09. Evaluate Extr M tv Monitorina proaram at TVA and particularly SON

{

I During the original investigation (reference A.11), NSRS found the l extremity monitoring program lacking and made the following recommendation:

The extremity monitoring program for SQN in particular and TVA'as a whole be evaluated for the capability of identifying extremity exposuro sources and measuring  !

and interpreting extremity exposures from all i

radiation sources encountered with emphasis on j seemingly point sources.

l The Radiological Health Staff responded that the program would be i evaluated in the second quarter of 1983 and a final report would be

}'

submitted to NSRS. For this follow-up review, NSRS reviewed the "Special Evaluation Report - Evaluation of Extremity and Multibadsing Dosimetry: Technical Review," part of the Radiological protection plan (Rpp), and various SQN documents. h special evaluation report

] was found to be thorough and addressed the major areas of concern j with the program, h recommendations made were appropriate. h l Rpp, revision 2, prescribes limits for extremity exposure and -

i requires measurement of extremity doses when the extremity dose j exceeds or is expected to exceed 25 percent of the quarterly limit, i

I

)i 29

The RPP also requires ALARA projob planning. RCI-10. " Minimizing Occupational Radiation Rxposure," requires a projob ALARA planning ,

report for several possible situations including the following; ,

1. " Handling of radioactive material Where extremity dose rates are in excess of 10 rem / hour at the working distance for the extremity."

2. "If an individual has received, from actual past exposures, greater than 25 percent of the extremity quarterly limit in one calendar day under one RWP."

RCI-3, " Personnel Monitoring," states:

3xtremity monitoring devices will be issued if an individual's extremity exposure rate exceeds the whole body exposure rate by a factor of five when the whole body exposure rate is greater than 100 mrom/he; when the whole body exposure rate is less than 100 mrom/hr, but the extemity exposure rate exceeds 500 meen/hr; or

• when the estimated extemity exposure is greater than 25% of the quarterly limit.

RCI-14. " Radiation Work Permit (RWP) Program " includes the following requirements:

1. Before issuing an RWP, the Health Physics representative must have a good working knowledge of the job location, equipment, radiological hazards which may exist in the area, and type of work to be accomplished.

2. Periodic radiological surveys will be performed in all areas covered by an active RWP.

3. Extremity and neutron sensitive dosimetry will be issued according to RCI-3 requirements. Individuals entering high radiation areas will comply with requirements listed in RCI-3.

Health Physics Section Instruction Letter HPSIL-1, " Radiation Surveys," prescribes how to conduct surveys and includes the following requirements:

1. Surveys will be made to determine dose rates, to detect any hasards to personnel that may exist, and to determine if the area is properly posted or barricaded. surveys.will also be made in potentially hazardous areas before entry by personnel.

30

= .

2. Equipment surveys will be performed on items to be worked on or to be shipped or moved to other areas within the plant.

3. Survey the area completely to locate any areas of higher dose rates that may be caused by equipment or penetrations.

4. Check the equipment completely to find the maximum dose rate.

5. If employees are to work on the equipment, determine the surface dose rate to estimate extremity dose. Estimate the surface dose rate by taking a E (window closed), WO (window open) reading.

Subtract the W from the WO and multiply the difference by the correction factor which was placed on the instrument during calibration. This is the beta dose rate. The W reading is the samma dose rate. Add the W and corrected WO for a total dose rate. Also, set a whole-body exposure rate by taking a reading at the closest point where the person's body will be.

Review of all these documents together indicates that a program is in place to identify extremity exposure hazards, preplan work for ALARA, and employ extremity monitoring devices when needed to ensure that extremity doses are accurately measured. This item is closed. .

N. I-84-12-SQN-01, Inadeouate Corrective Measures to Alleviate the t Dearaded CondiMon of the Thimble Tubes In the original investigation (reference A.25), NSRS recosmonded that responsibility for overall systems operability be formally assigned to plant engineers an4 those engineers be held accountable for periodically assessing the adequacy of the performance of the systems, the adequacy of instructions affacting the operation, maintenance or testing of the systems, and for assuring that problems are promptly identified and corrected in a quality manner. The responsible engineers should be required to keep inforined of industry and TVA information relating to the different aspects of the systems and to periodically forinally update plant management on the status of the system.

The Nuclear Power response in reference A.28 stated that while thimble tube blockage had existed and had been corrected several times during the life of SQN unit 1, it had never reached the post unit I cycle 2 blockage' level. The response also addressed system responsibility by stating:

The reactor engineering section has overall system  !

responsibility for the moveable detector system. This '

responsibility is recognized at the site. The reactor l engineering section is aware of and actively following l the proposed Westinghouse owners' Group program to '

address the thimble tube blockage problem, present assignments of " system Responsibility" are being reexamined as a consequence of the recent reorganization of the plant staff and site organisation. This reexamination will be an on-going process.

31

E The WSRS reply in reference A.29 stated that the response was * .

acceptable.

For this follow-up review, discussions were held with supervision and l engineers and documentation was reviewed. During discussions with "

] engineering supervisors, it was confirmed that the Reactor Engineering Section has overall system responsibility for the j moveable in-core detector system. A new procedure SQA 168, " System Engineering," was issued in January 1986 which assigned each plant f' system to a designated plant section (attachment A of SQA 168). The e Reactor Engineering Supervisor is responsible for System #94, "Incore

! Flux Detectors." Two reactoc engineers are currently assigned to this activity (one for each unit). They keep informed of industry jl and TVA infomation, including Westinghouse plant visits, Westinghouse Reactor Engineering Annual Seminar held in Pittsburgh, INFO, and NRC information notices, bulletins, etc. There is no i periodical update since the system status is reviewed on a continuing basis. This item is closed.

! 0. I-84-12-SQW-02, Inadeauste Survey and Feedback to Field Services a

Grous (FSG) Personnel

! In the original investigation (reference A.25), NSRS recommended that l in the future work assignments of this nature should be given to

those who are knowledgable of and will be responsible and accountable l for the success and safety of the operation to be accomplished. All

available information should be identifed and used.

The response in reference A.26 addressed reconenendations

, I-84-12-SQW-02 and -03 as one item. The following is an extract.

i l Sequoyah reactor engineering personnel contacted five '

nuclear plants questioning if they had cleaned thimble

! tubes at power and any problems they had experienced.

! The results of this survey were molded into the

overall plant decistorunaking process. The extent to ,

l which a survey of this nature should be carried out in

{ order to constitute an adequate survey is subjective j in nature. A survey is conducted only to establish an j adequate inforination base to facilitate management l decisions. In this case plant management felt that j they had adequate information to proceed with at-power 4

cleaning. SQNP believes the assignment of the survey to the reactor engineering section is consistent with their overall moveable detector system

! responsibility. While no survey can be all I encompassing, the additional information resources

! identified in the NSRS report have been noted for l future surveys.

l SQNP acknowledges that the personnel performing the f

! survey were not font 11er with the cleaning instruction i and had no experience with the actual cleaning operation. Again, the objective was to provide 32

i 1 * .

t j-l management with part of the information necessary to

make a decision regarding at-power cleaning. There l was no need for the survey personnel to interface ,

l directly with FSG personnel since FSG management i participated'in the discussions leading to the  !

ultimate decision to conduct at-power cleaning and

, were fully cognizant of eurysy results when making j subsequent work assignments, i

In retrosTect SQNP does not take issue with the fact I that the process used for cleaning the thimble tubes should not have been performed at power. SQWP l believes the decisionmaking process itself was sound

) even though weaknesses were evident in the '

j implementation process. ,

i l The NSRS reply in reference A.29 stated that the response, including

!, corrective actions, was acceptable.

For this follow-up review, discussions were held with supervision. ,

Based on discussions with supervisors and managers, it was determined l that there was no reason to doubt that work assignments would be j given to anyone other than those with the most knowledge and Who were available. Since the event, it has been stressed to everyone that 3 they are responsible and accountable for operations they are to {

accomplish. The staff also appears to be more sensitive to the l i possibility of other accidents. During preparation for the outage i i for unit 2, the revision 2 target schedule was developed. It is more z comprehensive than previous outage plans since it includes all items l 4

that could impact unit 2 start up. In addition to modifications, maintenance, engineering tests, operations, and design, it includes

{ SCR evaluations, NRC schedules, IE bulletine, employee concerns, NSRS  ;

i reports, environmental qualification program items, etc., These j items receive close management attention, particulsely if they are 1

likely to produce potential problems and delays. This also aids i

supervision in identifying knowledgeable individuals for work l assignments.

i I With respect to the action taken above, combined with the corrective

action taken for the other thimble tube report recommendations, it is  ;

concluded that the decisionmaking process and assignment of

knowledgable personnel to tasks has shown significant improvement.

4 This item is closed. 6

p. I-84-12-SQN-03, Inadeouste Decision Makinz Process  !

. In the original investigation (reference A.25), NSRS recommended that 1

for unique activities, plant management should take the time e necessary to identify and thoroughly evaluate hasards associated with ,

the activities using readily available inputs and obtaining '

information from knowledgable personnel who will be responsible and 1

accountable for the activity to be performed. Techniques such as a 1

33 i

i

systematic hasard analysis methodology to identify and derive an .

independent assessment of the hasards involved should be used.

The Nuclear Power response in reference A.28 stated:

Management meetings were held to discuss this activity and the potential hasards associated with it.

Discussions included the facts that (1) the work was to be performed on a pressurized system (2) any leakage from a thimble tube could not be isolated, and (3) there were radiological hasards associated with the work. h only weakness with this process may have been the lack of management involvement in the

. details f the work associated with the accomplishment of this maintenance activity. SQNP management is committed to ensuring future maintenance activity comply with normal plant practices. This includes procedure adherence, hazards and analysis planning (see our response to I-84-12-SQN-6), and encouraging input from those responsible and accountable for the maintenance activity.

N NSES reply in reference A.29 stated that the response, including corrective actions described, was acceptable. For this review'it was verified that the workplan hasard assessment worksheets were designed primarily as a tool to identify plant physical deficiencies and to determine the degree of hasard. Several hasard assessment worksheets were reviewed and appeared to be adequate for the intended purpose.

One request, dated November 29, 1985, from the Health and Safety Committee was for the Industrial Safety Staff to prepare an assessment on CO2 protected areas at SqN. Subsequently, the Industrial Safety Staff discussed the results of this assessment with the Committee.

It was also determihed that MONT' analysis has been used for job safety analysis, h Site Director needed to know if entry could be made safely into the unit 1 pnesuriser enclosure with the unit at full power. h object waa to open and close one root valve to the instrument line. h anslysis performed in November 1984 verified that the work could be performed safely and was subsequently implemented. h staffing'of the "new" maintenance planner positions also increases the effectivenessstf piraning unique work. This is described in detail in section IV.M (I-84-12-SQN-02) of this report.

h corrective action taken for this recomunendation combined with that taken for all of the other thimble tube report recommendations, as discussed in this report,' leads one to the conclusion that the decisionmaking process has been significiently strengthened. This is particulary true for those jobs. inolving safety and/or significant hasards. This item is closed.

34

.t l

• e 4

• I-84-12-SQN-04, Assianment of Work Function to the FSG as an Ordi m *F Q.

Work Activity i In the original investigation (reference A.25), NSRS recomumended that it should be emphasised to plant management that it is a fundamental i responsibility of management to assure that the knowledge and background of workers assigned to work functions is adequate and that l sufficient time and information be provided to properly plan and I

execute the work activity. The response in reference A.28 discussed the involvement of management and the plant sections in the 3

decisionmaking process and that it was not the intent of management

to create a sense of urgency to complete this job, but rather a

, responsible management decision was made that provided time to "

l demonstrate the success of the at-power cleaning technique. The fact '

i that a job normally perforined with the unit shutdown was being

} perforised at power may have produced an unncessary sense of urgency

} with the workers. In the future, the potential for this type of

mistaken worker perception will be eliminated by better cosuunication

between management and workers regarding operational schedules.

i In reference A.29 NSRS concurred with the SQN response. For this

follow-up review, it was deterinined that SQN has taken actions to emphasize management's responsibilities, that the knowledge and l background of workers assigned is adequate, and that sufficient time j and inforination are provided to properly plan and execute the work J activity. With respect to communication and time allotment a i

memorandum dated October 25, 1984, was sent to key SQN managers which

[ stated in part:

i

[

SQN management is comunitted to ensuring that future maintenance activities comply with normal plant l practices. This includes procedure adherence, hasards [

l and analysis planning, and encouraging input from

! those responsible and accountable for the maintenance I

activity. To eliminate the perception workers had

concerning thimble tube cleaning at-power'and any j future task that is performed at an unusual time, l better comununication will be established between '

management and workers regarding operational schedules and the urgency associated with them.

I j The managers acknowledged this memorandum by providing written i statements that they had meetings with their employees on the l

}

j memorandum. Additions 11y, the creation and staffing of the maintenance planner position discussed in section IV.H -

(I-82-21-SQN-02) of this report makes provision for better planning, and scheduling and safety hasard identification and analysis. The i assignment of the most knowledgable workers has been addressed in i

I section IV.0 (I-84-12-SQN-02). Additionally, SQN is undergoing, in the near future. INp0 accreditation for the crafts training program. l This program has the fundamental element of assigning only those j persons to perform tasks only if they have received specific training

} related to that task. This is discussed in section IV.LL 1

35

! l 3

J

.- _ ~ ._ . _ _ .

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

E

+

e l (R-85-02-SQN/WBN-01) of this report. The emphasis of the . .

i safety-first policy is discussed in section IV.J (I-82-21-SQW-05) of this report. Proper planning and execution will occur if it is done

well in advance of job performance. Review of the unit 2 schedule

! dated February 18, 1986, indicates that detailed planning and scheduling ic occurring.- Based upon the review of the above items, it is concluded that SQE is placing proper emphasis on assigning i knowledgable workers to jobs, performing detailed planning and scheduling, emphasising safety before schedule, and focusing responsibilties for proper planning and safe performance of jobs on j the appropsiate individuals. This item is closed.

i l R. I-84-12-SQN-05, Selection of an Insporopriate Instruction for the <

Control of the Work Activity In the criginal investigation (reference A.25), NSRS recommended that management should conduct an awareness program to reaffirm

! supervisor, engineer, and worker knowledge of the importance of j procedure controls, compliance with procedural requirements, and the

proper change process for inadequate procedures. Emphasizing the SQN policy as stated in SQA 129, which states that following instructions

! and taking the time to correct those which are inadequate are methods j to achieve nuclear safety.

l

The Nuclear Power response in reference A.28 addressed l

recosmondations 5, 7, 11, 17, and 22 as one item. The following is an extract:

In the future, a detailed scheduling process for incore thir.ble tube maintenance will be incorporated into the outage schedule and any deviations from scheduled work will be justified to plant management.

l . . . A problem existed in the coordination of the hold order and RWP associated with this maintenance activity. To alleviate this problem, Administrative Instruction AI-8 will be revised to clarify what moveable detector system maintenance requires a hold order and hold order requirements for RWPs will be modified to indicate AI-8 will be followed.

NSRS replied in reference A.29 that the response,' including corrective actions, was acceptable.

A memorandum (reference A.33) was sent on October 25, 1984, to plant management listing items that managers needed to stress to their personnel on a continuing basis. Included was the affirmation of coupliance with procedures and management policy being safety first.

The safety-first aspects are discussed in detail in section IV.J,  ;

(I-82-21-SQN-05) and section IV.GG, (I-84-12-SQN-20) of this report.

Better communication was to te established between management and workers.regarding operational schedules and the urgency associated with them. The managers responded when they implemented the policies in the reference A.33 memorandum.

36

. = .

Section 1 of SQA 129 states that: ". . . . following instructions j and taking the time to correct those which are inadequate are methods to achieve nuclear safety". At the 8:15 a.m. Plant Managers daily meeting, the question is asked whether there were any failures to  ;

)_ follow procedures. This emphasizes procedures controls. Also, it was determined that: (1) the current outage schedule details tasks

{ for the incore thimble tube, (2) AI-8, Access to Containment," has  ;

! been modified to clarify hold order clearance and EWP issuance for  :

lower containment related to the incore flux drive, as discussed in I

section IV.U. (I-84-12-SQN-08) of this report, (3) the RWP coversheet j has been modified to state that " Entry into containment will be ,

performed in accordance with AI-8.", as discussed in section IV.U (I-84-12-SQN-08) of this report, and (4) the emphasis on the change control process for inadequate procedures is being satisfactorily

~

pursued, since training on the use of plant instruction change forms has been provided as discussed in detail in section IV.II, (I-84-12-SQN-22) of this report.

It is concluded that the corrective actions described above satisfactorily emphasize procedure compliance and change control and the safety policy. This item is closed.

S. I-84-12-SQN-06, Inadeouste Job Safety Analysis and Hazards Assessment In the original investigation (reference A.25), NSRS recommended that

! the job safety analysis program be upgraded. An effective hazards

assessment methodology should be established as a tool to be used to j' analyze the identified radiological and industrial aspects of the job; the probability of an accident; and the impact on the workers,

! plant, and the public. Additionally, implement the recommendations j of NSRS Report No. I-82-21-SQN.

The Nuclear Power response in reference A.28 stated that both a job safety analysis and a work place hazard assessment methodology are in place for evaluating, preventing, and/or mitigating accidents at i

SQNP. The office of Nuclear Power will continue to examine the

existing workplace hazard assessment methodology to determine its

applicability as a tool in job safety analysis. The Industrial i

Safety Engineering Section subsequently (reference A.32) determined ,

that the workplace hazard assessment methodology cannot be used effectively in job safety planning,- but it is effective in qualifying and prioritizing physical deficiencies. The reference A.32 also referenced the job safety planning procedures used at SQN, Which is considered to be appropriate.

For this follow-up review it was determined that the recommendations of NSRS report I-82-21-SQN have been implemented. Sections IV. H, I,

J, K, L, and M (I-82-21-SQN-2, -4, -5, -6, and -7, respectively) of l

this report documents the closure of the items-that were open from this report at the beginning of the follow-up review. ALABA preplanning criteria is incorporated in RCI-10 " Minimizing occupational Radiation Exposures," as discussed in section IV. Z (I-84-12-SQN-13).

37

The safety aspects of the job has been improved by the newly created.

j and staffed positions of maintenance planner, whosa job description o includes safety responsibilities as described in section IV. H (I-82-21-SQN-02) of this report.

Prior to the thimble tube event there were two safety groups--plant and field services. .These were combined in August 1984. This l consolidation of safety skills should result in improved safety reviews and assessments.

More detailed analysis and planning of unusual activities was verified as being conducted in the review of the pressurizer enclosure entry with the unit at full power hazard analysis which is described in section IV. P (I-84-12-SQN-3) of this report.

i Although probabilities of accidents are not quantified in the job safety analysis and/or hazards assessment methodologies, conservative failure / effects assumptions adequately address failure probabilities and potential impacts on workers and the public.

The incorporation of extensive ALARA preplan requirements detailing radiological hazards into RCI-10; creation and staffing of a maintenance planner position with safety responsibility; performance of hazard assessments; the existence and use of procedures for job safety planning, the combination of two plant safety groups; the use of conservative failure modes in hazard assessments; and satisfactory corrective action being taken on all of the NSRS recommendations from '

the I-82-21-SQN report, satisfactorily resolves this item. This item is closed.

T. I-84-12-SQN-07. Inadeouste Field Quality Enaineerina (F0E) Review of Maintenance Request (MR) and Reference Work Instruction

, In the original investigation, NSRS recommended that SQ'- .'ould

( improve the quality of the FQE review process of MRs to assure the i quality of the referenced work instructions, the proper program controls are used, and the instructions are appropriate for the activity being performed.

The Nuclear Power response in reference A.28 addressed recommendations 5, 7, 11, 17, and 22 as one item. The following is an extract:

The MR was reviewed by FQE as part of their responsibility to ensure an adequate procedure exists for the performance of the work. A job safety analysis was performed by the maintenance foreman as required by the MR process.

After thoroughly analyzing this event and the NSRS conclusions, SQN acknowledged that the MR and FQE's review of the MR did not meet the

! requirements of the Sequoyah standard practice en maintenance j management. SQN 2. SQN was to review the MR system and QA review process to ensure no programmatic deficiencies existed. NSRS replied in reference A.29 that the Nuclear Power response, including corrective actions described, was acceptable.

38

- _ . _ - . . .= - . . - - _ - . ._ --

For this follow-up review, it was determined that an evaluation of the NR process was conducted by the Quality Engineering Group. The evaluation indicated the need for training personnel involved in the

. MR planning process and to have more supervisory involvement in the j process. Adjustments were made to upgrade the QA review program.

j Also, the recent training records for QA reviewers were examined.

Initial MR review training was conducted January 30, 1986, and final i NR review conducted February 6, 1986. A review of the QE Section Instruction Letter (SIL 5.3) in effect during 1984 shows it to have l been adequate. The rejection rate of final NRs indicates a thorough check of the MRs, but also indicates a need for improvement of MR preparation. Sections IV.R. I, DD, and II (I-84-12-SQN-5, -11, -17, and -22, respectively) of this report address other satisfactory corrective actions related.to procedure adequacy.-reviews, and controls. This item is closed.

j U. 1-84-12-SQN-08, Noncompliance with Requirements of RWP No. 01-1-00102 In the original investigation (reference A.25), NSRS recommended that it be emphasized to plant employees that compliance with the requirements of RWPs is essential for their own protection.

, Nuclear Power essponded in reference A.28 that the noncompliance

! resulted from confusion existing in AI-8, " Access to Containment,"

with respect to hold order removal. The response also stated that special instructions for RWPs would be modified to indicate AI-8

! requirements were to be followed and AI-8 would be revised to remove I

the confusing instructions on hold order removal. NSRS, in reference A.29 concurred with the acceptability of the Nuclear Power action.

1 For this follow-up review, it was determined that the NWP cover sheet I and AI-8 were revised to eliminate potential confusion, and the RWP

] coversheet (form TVA 7903B (DNP-9-84)}, has been revised to add an l item 5 that states: " Entry into containment will be performed in accordance with AI-8".

i AI-8, section 2.4, has been revised to clarify hold ordar removal as

follows

Prior to entry into lower containment or the annulus the incore flux detectors shall be verified to be in the storage position or inserted to within ten (10) feet of the core. The SE shall initiate a hold order clearance on the incore flux drive motors control power. This hold order shall remain in affect until the SE is assured all personnel have beer. cleared from containment and the personnel access is locked. Prior to issuing a radiation work permit (RWP) for lower containment or annulus, HP shall verify incore j detector system is tagged with a hold order. The SE i

39 a

r i

f

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

+

will issue incore detector system hold order clearance , ,

to HP Shift Supervisor by title. The incore detector hold order clearance will remain issued to HP Shift

, Supervisor by title at all times except when running core maps or performing incore detector system to be operated while persons are in the incore instrument ,

room. I i

Maintenance work that requires the incore detector I system to be operated while persons are inside the incore instrument room shall be coordinated by i Operations HP, and the applicable maintenance section. (Other work in progress in lower compartment or annulus will be evaluated for continuation by this group.)

The SE will contact HP to verify no one inside lower containment or annulus and Public Safety to verify access is locked prior to releasing clearance on incore detector system except as outlined above.

3 The modifications to the RWP coversheet and AI-8, combined with the i

following, adequately resolve this recommendation: emphasis on compliance with RWPs is provided in GET training; RWP and ALARA ,

training has been given to a significant number of plant personnel as described in section IV.Z (I-84-12-SQN-13) of this report; personnel safety issues have been emphasized as discussed in sections IV.I S, ,

Z, and GG (I-84-12-SQW-04 -06. -13. -20, respectively) of this ,

report; adherence to hold order clearance procedures training has been provided to those personnel authorized to receive clearances as ,

discussed in section IV.V (I-84-12-SQN-09) of this report. This item  !

is closed.

V. I-84-12-SQW-09, Woncompliance with Requirements of Section 5.1.4 of AI-3. " Clearance Procedures" In the original investigation (reference A.25), NSRS recommended that since the hold order system is the method used at SQN for the

• protection of workers, the public, and equipment, strict compliance ,

with the requirements of AI-3 should be emphasized and enforced. NUC >

PR responded in reference A.28 by stating that additional emphasis will be placed on making all personnel aware of the requirements for l

the person responsible for work to be on the clearance. -This will be accomplished in preoutage briefings, existing clearance procedure training classes, and the periodic management safety meetings which are attended by managers, foremen, and engineering personnel. In reference A.29 NSRS concurred with this corrective action.

For this follow-up review, discussions were held with operations training personnel and a review of training documentation was

-performed. SQW management determined that a total retraining to AI-3, " Clearance Procedures," was necessary for those personnel authorized to receive clearances, as specified in Appendix A of j AI-3. This retraining was performed in the timeframe of Apell I

40 i

+ .

through September 1985 to the AI-3 lesson plan (reference I.12). The three hour retraining contained a formal written examination. A grade of a least 80 percent is passing. At the completion of the retraining program, the clearance authorization list of Appendix.A to AI-3 was revised (December 16, 1985) to reflect only those personnel identified by management and had successfully completed the retraining were to be placed on the authorized list. A random selection of three of the persons on the Appendix A list of December 16, 1985, and their examination results revealed that these individuals had achieved a passing grade on the clearance procedure retraining examination. The Operations Section did not take this specific AI retraining since the licensed personnel receive clearance procedure training in their requalification effort.

The instructions AI-3 and AI-8, " Access to Containment," have been modified to clarify the hold order initiation and removal requirements. Section 4.3 of AI-3 states:

The SE shall, issue the hold order clearance on the incore detector systems to health physics shift supervisors by title, per AI-8 (both units).

] Section 2.4 of AI-8 " Access to containment," has been clarified with respect to issuance of an RWP and hold orders as discussed in section

! IV. U (I-84-12-SQN-08) of this report. Additional evidence of hold order discussions was determined to have taken place as follows:

1. The SNP unit I cycle 2 outage critique meeting minutes documents management concerns on hold order usage and recomunends

retraining personnel on AI-3. As stated previously, this has been accomplished. Thus, there is evidence that hold order clearance problems are reviewed in outage briefings.

2. Employee crew safety meeting reports (SQS 7) from November 1985 through January 1986 were reviewed and it was determined that en electrical day crew and machinist midnight crew had held 4

discussions on hold orders.

1 4

3. The planning supervisor stated that the planners are instructed

• to minimize the number of hold orders.

a SQN has adequately addressed the recommendation by revising procedures, retraining personnel on clearance procedures, discussing hold order clearances in periodic safety meetings and outage critique j meetings, and minimizing the use of hold orders. This item is closed, t

W. I-84-12-SQN-10, Modification of Cleaninz Tool Base Suocorts Without Performina a Technical Evaluation or Testina 3 In the original investigation (references A.25), NSRS recoseended' that it be emphasized to the plant staff that changes to tools and equipment affecting work on critical structures, systems and components (CSSC) can be made only after a thorough technical 41

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

l l

evaluation has been made on the effect it will have on the system and.

used only after the modified tool or equipment has tested satisfactorily. Nuclear Power responded in (reference A.28) that SQN l will review "special tools" and evaluate the need for modification controls for these types of tools. In reference A.29, NSRS concurred with the Nuclear Power response.

For the follow-up review, it was determined that a SNP Standard Practice SQM-63, "Special or Modified Tooling-Primary Systems," has been prepared and is in use at SQN. The stated purpose of the instruction is:

.. .to provide a means to evaluate special or modified tooling that is used in conjunction with maintenance or modification activities which could directly or consequently cause adverse effects to primary systems.

The requirements of this practice are applicable to special or modified tooling which could directly or consequently cause adverse effects to primary systems. The instruction specifically states-l In general, special tools used on equipment fitting the following criteria fall within the scope of this instruction. j 2.1.1 Components Which are in service, pressurized or energized.

l 2.1.2 Components which, if the tool caused failure of the component, could cause loss of primary coolant or the loss of uncontrollable amounts of radioactivity contaminated water during the use of the tool.

l 2.1.3 Components Which, if the tool caused failure of the component, could cause the loss of l safety function while the tool is being used.

It was detenmined that the SQN personnel have knowledge of the existence of this standard practice and that it is being actively used for special tool evalustions for ccmponents other than those described above, e.g. a special tool evaluation for a motor lifting eye. Although the instruction does not specifically identify CSSC within the scope, the items 2.1.1, 2.1.2, and 2.1.3 above combined with the knowledge of and evidence of.its use in a general manner, satisfies the intent of this recommendation.

The testing aspect of the special tooling and/or modification thereof is covered in section 5.4.b of SQN 63 Which states:

If the engineer determines that an evaluation is needed, he performs the evaluation based on the following criteria:

a. Plant condition while tool is being used (i.e.,

system pressurized, s Mode, draine d, and unit at power,ystem etc.) isolated and 42

. .-. . - - - - . . . - -- _.. = --- --

j . .

i *

b. Testing, if feasible (such as mock-up of the tool as it is proposed to be used).

c. Technical evaluation (such as a stress analysis of the tool when used as proposed).

d. Logical assessment of the tool using engineering principles to deduce the effects of the tool.

e. Vendor recommendations, if the tool was procured for a special task.

Testing is specified as one way to evaluate the tool and the alternatives specified are considered to be acceptable.to demonstrate adequacy, thus satisfying the intent of the reconumendation.

Based on discussions with the mechanical maintenance engineering section supervisor, 100 percent of all their Maintenance Instructions were reviewed and special tooling was identified. The special tools

) were evaluated according to SQ4 63. Conversations with the electrical maintenance engineering supervisor verify that he has a working knowledge of SQN 63. This section recently completed a special tool evaluation per SQM 63 for a containment air fan motor lifting eye. A review of twenty special tool evaluations verifies that the SQN personnel are using the special procedure in the daily operations. This item is closed.

I. I-84-12-SQN-11, Vloistion of Work Instruction 4 In the original investigation (references A.25), NSRS recommended that management should emphasize to the plant staff that adherence to PORC reviewed, plant manager approved plant instructions is mandatory and a requirement of the Technical Specifications and that instructions and controls established to assure nuclear and industrial safety. Periodic assessments of compliance with instructions should be initiated and corrective actions taken to correct weaknesses observed.

Nuclear Power replied in reference A.28 stating:

SQNP did not believe generic program weaknesses have been indicated by this event. However, SQNP management understands I

their detailed involvement in how the job was to be implemented j

during the evaluation to determine its feasibility may have {

1 unintentionally sent a message to key implementing employees l 1

creating the. impression they had authority to proceed without '

adherence to normal plant practices.

I For this follow-up review, it was determined that in the staff i

i meetings held at 8:15 a.m. each morning at SQN, the Plant Manager questions whether there were any failures to follow procedures. In this way, the requirements for adherence to PORC-reviewed, plant manager approved plant instructions is made perfectly clear.

43 i

__ _____ _ . _ . - - - - - - - - - -- - -- ~ - ' ~ ^ ' '~ ~

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

l t

A review was made of the 1985 Quality Surveillance Section Annual ,

Plan. Periodic assessments of compliance with instructions had been conducted through December 1985. All plant activities with the i exception of site emergency plans had been conducted. Additional corrective actions on quality of and compliance with procedures is i discussed in section IV. GG (I-84-12-SQN-20) and IV.GG (I-84-12-SQN-22) of this report. This item is closed.

l Y. I-84-12-SQN-12 Lack of Control of Earess Capability from Containment i

l In the original investigation (reference A.25), NSRS recommended that a policy and methodology be established requiring an evaluation of the effect on work in progress and notification of affected workers as necessary before granting permission to incapacitate egress routes from the reactor building containment. Emphasize to plant managers and workers that working in the reactor building containment involves some risks end controls for containment integrity are established.

Identify the risks involved and established controls to the employees.

i In (reference A.28) Nuclear Power responded: (1) that the submarine hatch was nearby and available as an unhindered egress route, (2) agreed that the reactor building egress should not be impaired when maintenance or other activities within containment are necessary while the unit is at power conditions, (3) good communications are essential and policies regarding communication were being reviewed to ensure effectiveness while maintaining flexibility for the shift engineer to evaluate such situations on an individual basis and l determine the extent of notification required, and'(4) SQN acknowledged that the FSG personnel were not adequately aware of the Technical Specification requirements associated with the containment airlocks and that future emphasis would be placed on ensuring responsible maintenance personnel are made aware of the Technical Specifications associated with the airlocks on a job-by-job basis.

NSRS responded to the Nuclear Power response in (reference A.29) generally concurring with the position that workers are entitled to l

know of any condition which could impair their egress from

, containment should rapid egress be necessary. This is fully l consistent with TVA's safety-first policy. Although the Nuclear Power response commits to a review of present policies regarding communications, NSRS stated it would be looking for specific actions taken following this review activity, i The NRC investigation of the thimble tube event stated that the failure to establish guidance or positive controls in AI-8 to the

! operations staff for changes in airlock access status during Modes 1 l through 4 was a further example of TS 6.8.1 violation.

44 I

l l

For this follow-up review', it was determined that AI-8, " Access to 4

containment," has been modified to address egress capability When airlock doors are made intentionally inoperable. Section 2.6 of AI-8

. was added and states:

The upper and/or lower containment airlock doors shall not be intentionally made inoperable (pervent personnel egress) while personnel are inside containment. .If the doors must be made inoperable with personnel inside, they will be instructed to use an alternate exit.

Additionally, section 2.4 of AI-8 requires that a survey of the lower containment and annulus be made for personnel presence prior to L releasing the clearance on the incore detector system, as discussed i in section IV. U (I-84-12-SQN-8) of this report.

J Those portions of the recommendation related to emphasizing to plant managers and workers of risks involved is addressed in GET training and work instructions. Specific on-the-job safety concerns are s addressed further in the details of sections IV. H J. S. U, V, 2 and GG (I-82-12-SQN-2, -5 and I-84-12-SQN-6, -8, -9, -13. -20) of this report. This item is closed.

Z. I-84-12-SQN-13, Breakdown in the ALARA Preplanning Prozesa '

, In the original investigation (reference A.25), NSRS recommended that it be emphasized to the plant staff that compliance with ALARA peoplanning requirements as specified in RCI-10 must be accomplished.

2 Nuclear Power responded to this recommendation in reference A.28.

The response stated:

SQWP supports and practices ALARA preplanning based on expected doses with consideration given to potential doses. In concert with corporate policy, it is the plant's goal to maintain radiation doses ALARA in all our work activities. . . .Since the time of the thimble tube ejection incident.

l RCI-10 has been revised to include specific ALARA preplanning criteria.

The NSRS response in reference A.29 stated that further discussions

in the area of ALARA preplanning would be required, i

For this follow-up review, additional documents were reviewed and discussions were held with Health Physics personnel. It has been determined that significant positive actions have been taken and are ongoing to improve the ALARA preplanning program. These are: (1) modification of RCI-10. (2) ALARA/RCI-10 training has been for some operations, HP, mechanical maintenance personnel, and instrumentation maintenance weekly safety meeting, and (3) participation in PORC subcommittee biennial review of existing plant instructions and for

, new instructions. The following provides a brief discussion related to these efforts.

! 45

,,,..-._-,,.,.,,m_, , , , , -umv, ,,_m._m

l

1. Modification of RCI-10 , ,

i Extensive RCI-10. " Minimizing Occupational Radiation Exposures,"

l revisions have taken place since the thimble tube event. Before the event, the procedure required an ALARA preplanning report to be completed by the responsible supervisor if the job had a potential exposure greater than 5 man-rem. The instruction has been revised to require an ALARA preplan report to be filled out by the responsible supervisor when in his estimation any specific job meets any of the following conditions:

o If an individual has received, from actual past exposures, greater than one rem (whole body) in I one calendar day under one radiation work permit (RWP).

o If an individual has received, from actual past exposures, greater than 25 percent of the extremity quarterly limit in one calender day under one RWP.

o If the work area dose rate exceeds one rem / hour (whole body).

l o Handling of radioactive material where extremity j dose rates are in excess of 10 rem / hour at the working distance for the extremity, o If an individual is expected to receive greater than 10 MPC hours in one day (after appropriat protection factors are applied). ,

o If an individual is expected to receivi greater than 40 MPC hours in one week (after appropriate protection factors are applied).

o The collective dose is expected to exceed 5 man-rem under the RWP for each individua1 job.

o All jobs shall be preplanned when deemed necessary for exposure control by the ALARA engineer.

In addition, the preplanning and postplanning checklist was expanded to 41 items and approval requirements were expanded from the job planner and ALARA coordinator to the current level of preparer, cognizant engineer Health Physics Shift Supervisor, section supervisor and ALARA engineer.

These changes are considered to be significant improvements to RCI-10 and the preplanning effort. Current modifications to the preplanning checklist simplification are being considered as are

! positive modifications to the RWP instruction RCI-14. " Radiation Work Permit (RWP) program".

46

o. .

The ALARA engineer stated that 104 preplans were prepared in 1985'and 12 in 1986 as of February 21, 1986.

2. Conduct A14EA/RCI-10 Trainina

! The ALARA engineer records show that he provided training instruction to 237 SQN personnel from the Operations, Mechanical Maintenance, and Instrumentation Mechanics Sections. These were conducted in combinations of one-hour lectures with handouts.

i Several MPC hour tracking training sessions were also conducted at weekly instrumentation safety meeetings.

3. Plant Instruction Reviews 4

j The HP Group is represented on the PORC subcommittee's biennial review of current plant instructions and all new ones. If deemed applicable a precaution statement is inserted for ALARA and RWPs. Section 4.1 of SMI-0-68-28," Change Out of the RCS j Narrow Range RTDs" is an example of the statement being inserted j into applicable procedures. This section states:

Contact HP for applicable radiation work permits (RWP) AI-33 shielding and ALARA preplanning.

This insertion is an improvement, since it instructs the personnel conducting the instruction to verify that ALARA preplanning has been performed, if necessary. At the time of this review, the HP technical supervisor stated that approximately one-third of the

, potentially affected instructions had been reviewed and that by 1988

! all currently existing procedures will have been reviewed for RWP and l ALARA preplan precaution determination.

}

f Based upon the revision to RCI-10, the conduct of the training, and j the plant instruction review, it has been and is being emphasized to

, the plant staff that compliance with ALARA preplanning requirements

! as specified in RCI-10 is required. This item is closed, i

i AA. I-84-12-SQN-14. Need for Formal Documentation for Upper Plant l Manatement Approval to Work in Radiation Dose Rate Fields Greater 4

Than 50 Rem / Hour 1

In the original investigation (reference A.25), NSRS recosmonded that SQN establish formal requirements and a method to document-

! authorization to work in dose-rate fields greater than 50 rem / hour. l

{ Wuclear Power responded in reference A.28 by stating that at the time l of the thimble tube event, RCI-14. " Radiation Work Permit (RWP)

! Program, section IV.B.6, specified that the Plant Manager.was required to review the RWP When dose rates exceed 50 rea/ hour; the

, appropriate management personnel were notified and verbal .

authorization given to' continue the job; RCI-14 was revised requiring formal documentation of this' review and authorization, and the

! appropriate RWP signature sheets were being revised to include a l signature slot for the Plant Manager, if required.

47 i

. - . , , - . - , , , - - - - - , ,, ..--,_,r,,.,,. ,.w. . , . _ . - , ,...,,,m, - - , . . , , ..m._, . - ~ ~ . . . , , . , .w n ,m.,.,.m.,m,,.,,-,,eo.m,,., %r_.m . , , . , , , , , ,-

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

For this follow-up review, RCI-14, revision 4, dated July 10 1985,.

was examined. The requirements for Plant Manager review and approval are stated in sections IV.B.7 and V1.F. These state that the Plant Manager shall review and approve all RWPs when the work area dose l1 rate equals or exceeds 50 res/ hour or prior to any entry inside the l

polar crane wall when the reactor is at power. Also, the Plant Manager will indicate any additional special instructions to be followed on the RWP. The RWP provides a signature slot for the Plant i Manager. This item is closed.

BB. I-84-12-SQN-15, Availability of Communications Following the Accident In the original investigation (reference A.25), NSRS recommended that anytime the telephone is out of service in the airlock, alternate i communication methods should be considered and employed.

Additionally, availability of communications should be considered during the perforumnce of the job safety analysis and job planning.

NUC PR responded in reference A.28 that SQN acknowledges that the 1 airlock telephone was inoperabia and that additional emphasis would i be placed on timely response for maintenance requests on these phones.

Section 6.6 of the NRC investigation report of the thimble tube event stated that While regulations do not address airlock communications,

' the potential for airlock operating mechanism failures and for events inside containment appears to justify maintaining reliable conununications between airlocks and manned stations outside containment.

4 For this follow-up review, the NSES reviewed AI-8 " Assess to containment," revisions 10 through 16, and SQM 2. " Maintenance Management System," revision 16. At the time of the thimble tube event AI-8 did not specifically require verification of the

(, operability of the containment airlock telephone. Subsequent revisions to the procedure were made to add the requirement to check the airlock phone upon entering. Section 6.0 of the current AI-8, revision 16, requires the Public Safety Officer unlocking the airlock to ensure that the phone inside the airlock is checked for proper  :

operation prior to the first person entering containment and to fill out a data sheet related to phone operability. If it is discovered that the phone is not operating properly, Public Safety shall initiate an immediate attention list MR to electrical maintenance for repair. Entry into containment during the period the phone is out of service shall be approved by the individual's supervisor or the shift engineer (SE) when the supervisor is not present.

The SQM 2 states that inunediate attention WRs may be commenced at anytime, but normally would not bump a job already in progress. It i should, however, be started within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is considered satisfactory that during the 24 plus hours that the phone could be out of service entrance is controlled by the supervisor or SE.

48 l

I

i l

- . - . .__..--.-..---_-.,...----m.--------.--- - . - - - , - - - - _ _ . - . - _ , , - - - - - - - - - , - . , ---_ -._,---.--.,--- . - -

1

I

. . l i 1 I ,

- i

!~ l With respect to availability of communications being considered during performance of the job safety analysis and job planning, the

following action was verified to have been performed

(a) RCI-10, l " Minimizing Occupations 1 Radiation Exposure," was revised to add a prejob ALARA planning report checklist which requires a determination of whether special communications equipment is needed to enable workers to communicate effectively while in radiation areas, (b) the  ;

job hazards analysis that was performed for entry into pressurizer

enclosure with the unit at 100 percent power specified alternate '

egress routes and required, as a form of commamication, posting an j employee inside upper containment at the airlock wheel to open it in

?

the event of an emergency, and (c) AI-8 was modified to require

} notification of personnel in containment of alternate egress routes

in the event the airlock door is intentionally made inoperable. The i combination of all of these actions demonstrate that comununications are being considered as a part of the job planning and hazard j assessment efforts. This item is closed.

j CC. I-84-12-SQN-16. Effective Cleanina of the Thimble Tubes by WUS 1 Corporation 1

In the original investigation (reference A.25), NSRS recommended that j WBN be advised of the effectiveness of the NUS cleaning method over j the Teleflex method. Nuclear Power responded in reference A.28 that ,

1 the NUS thimble tube cleaning method appears to be effective and that

) they will advise WBN of the NUS technique. The. response also stated

! that the ultimate effectiveness can only be judged af ter considerably j more operating experience. WSRS concurred with the response in i reference A.29.

For this follow-up review, it was determined the SQN plant Manager l sent a TVA 45D to the WBN Plant Manager informing him that the NUS j thimble tube cleaning method appears to be an effective means to

! clean the tubes. The WBN mechnical maintenance personnel stated that-l the NUS thimble tube cleaning system has been purchased by WBN. The j

WBN cleaning procedure MI-94.3, "Incore Flux Thimble Cleaning and Lubrication," section 3.5 states: "This instruction is to be performed using the NUS supplied flux thimble cleaning equipment."

This item is closed.

DD. I-84-12-SQW-17, Poor Quality Cleanina Procedures and Ie%uate PORC

Review

! In the original investigation (reference A.25), NSRS recomunended that l SQN should evaluate the PORC review process and consider .)

! supplementing the review process with expert subcomunittees to  ;

! properly evaluate procedures and advise the Plant Manager on their j l

adequacy before he approves or disapproves. Additionally, cancel

SMI-0-94-1 and do not use SMI-0-94-2 again until it~has been revised i to include at least the quality elements listed above. perform a

! generic review of all maintenance and special maintenance f instructions to ensure adequacy.

t The response in reference A.28 addressed recommendations 5, 7, 11, I

17, and 22 as one item. The following is an extract.

I

\

! 49 i

L-._._.- _ -. .-._..-.__ _ _ . - .._ . . _ . , . _ _ _ . . _ _ _ _ _ _ . _ _ . _ _ _ . _ ._.-~_, _ _ _ . ~ _

i' After thoroughly analyzing this event and the NSRS . .

! conclusions, SQWP acknowledges the following: (1) The

' work package (SMI-0-94-1 and MR) provided poor quality i instructions in that they were not revised to reflect i

! at-power cleaning and did not meet technical )

specification requirements for this maintenance -

activity. This procedure has been cancelled. (2) ,

, SMI-0-94-2 did not contain all the quality elements i necessary for this maintenance activity and it is i being revised to reference Maintenance Instruction MI-1.9 "Botton Mounted Instrument Thimble Tube l Retraction and. Reinsertion" for the disassembly and assembly of the 10-path transfer devices. Appropriate

! cautions and warnings are being added to prevent l' damage to the mechanical seals. Postmaintenance inspections and testing requirements will be added to i SMI-0-94-2; however, it should be noted that this l procedure previously contained a double signoff that precluded its use at power.

l SQNP does not believe generic program weaknesses have

been indicated by this event. However, SQNP l management understands their detailed involvement in how the job was to be implemented during the evalulation to determine its feasibility may have unintentionally sont a message to key implementing employees creating the impression they had authority

, to proceed without adherence to normal plant practices.

I

! NSRS replied in reference A.29 that the response, including

corrective actions described, were acceptable. For this follow-up

! review, discussions with the Plant Manager showed that an evaluation '

of the PORC procedure process had been conducted. It had been concluded that the work load in PORC needed to be reduced. The review process was optimized by identifying Who the procedure was to

be routed to, and minimizing the number of reviewers. The current thinking is not for expert subcommittees but to get the procedure review out of the PORC process, by replacement with a qualified procedure reviewer. This would necessitate a Technical Specification change and would be unlikely to occur very soon. It appeared that

. Carolina Power and Light had obtained approval from NRC for such a change and the Plant Manager was following this up. The adequacy of

procedure reviews is discussed in more detail in section IV.II

{ (I-84-12-SQN-22) of this report.

! Procedure SMI-0-94-1 was cancelled on October 9, 1984, and SMI-0-94-2 cancelled December 11, 1984. MI-1.10. "Incore Flux Thimble Cleaning

and Lubrication," was approved as a thimble tube cleaning procedure i on October 31, 1984 with a current revision date of September 9, i 1985. This procedure and SWAGELOK fittings is discussed in Section IV. MM (P-85-02-SQN/WBN-02) of this report. Generic weaknesses in l maintenance instructions and ongoing corrective action are discussed l in Section IV. II (I-84-12-SQN-22) of this report. This item is closed.

i 50

j + .

Ed. I-84-12-SQN-18, Noncompliance with Serious Accident Reportina and Accident Scene Preservation Requirements f

f- In the original investigation (reference A.25), NSRS recommended that SQN determine the cause of the noncompliance and take corrective

{ actions as necessary to ensure future compliance with established requirements.

The Nuclear Power responded in reference A.28 stating:

1 This event was initially considered in terms of its radiological impact with recovery to reduce exposure as its optimum concern. Industrial and radiological safety were both considered during this recovery.

Approximately 12 days after the event a team from NUC j PR was designated to review the industrial safety aspects of the accident to determine if it fell under

the TVA Serious Accident Investination Procedure and, if not, to proceed with a report highlighting lessons learned. The Designated Agency Health & Safety

( Official (DASHO) and the Manager, Office of Power, were notified at this time. The team concluded that this event did not meet the requirements of the agency's procedure and made that recommendation to NUC PR management.

1 .

The Office of Nuclear Power acknowledges the need to review existing TVA reporting and investigation

) requirements for industrial safety incidents and,

Where needed, will provide clarification on When these j requirements are applicable. This review will also focus on defining requirements related to the nuclear safety and radiological aspects of an incident and. '

, should be complete by January 1, 1985.

i With regard to the NSRS concern on preservation of the l accident scene, the accident scene immediately after

the event' was extensively recorded by photographs. In-any event, it would not have been possible for either a division-level or agency-level team to actively l

investigate the scene of the accident due to the high postevent radiation fields present in the incore i instrument room.

The NSRS response in reference A.29 concluded that

)

i This event fell within the guidelines of the TVA l Serious Accident Investimation Procedure based on the amount of damage involved (including cleanup costs as clarified by NUC PR personnel). It appears that the existing criteria is not specific enough to identify.

when events should be investigated by an independent j organization. We concur that existing procedures for 51

t reporting and investigation of accidents / incidents - .

need to be reviewed and revised to address events of a nuclear safety / radiological nature. As a part of our follow-up effort we will examine changes you make to the existing corporate and plant procedures to address these concerns.

l The Nuclear Power response in reference A.36 stated that:

I The Office of Nuclear Power is developing a procedure to address investigations of both industrial nuclear safety and radiological incidents. The composition of the investigative teams will be addressed in this

procedure and will, of course, be dependent upon the t

nature of the incident. In the. interim, we feel we demonstrated during the recent Browns Ferry unit 3 startup the ability to designate an investigative team with the appropriate expertice.

j For this follow-up review, it was determined that the Nuclear Power procedure for Serious Incident Investigations was issued May 14, 1985 (reference A.47). The revisions to the site procedure for Accident  :

Reporting and Investigation SQS 29 were incorporated and approved July 18, 1985.Section IV. FF (1-14-12-SQW-19) discusses the content of SQS 29 and it is concluded that the procedure adequately addresses team somber independence, preservation of the scene of the accident, timeliness and content of ths investigation and repo:* This item is-closed.

l l FF. I-84-12-SQN-19, Limited WUC PR Accident Investimatien l

In the original investigation (reference A.2$), NSRS recommended that during future accident investigations appropriate personnel should be I appointed to eliminate any potential conflict of interest; the i investigation should be initiated as soon as possible after the accident as prescribed by established procedures; sufficient time should be allowed for conduct of the investigation; and it should encompass all aspects of the accident including programmatic weaknesses or breakdowns, and nuclear and radiological safety. The Nuclear Power report should be revised to delete the recommendation that consideration should be given to leaving the inner door open during such activities.

The Nuclear Power response in reference A.26 stated:

The investigation team was named to perform a specific function as stated in finding I-84-12-SQN-19. If it had determined that a serious potential did exist, the agency level team (AIT) would have been named by the DASHO and office manager. The division level teams would at that time have been disolved. In all probability the SQNP FSG supervisor would not have been designated to serve on the AIT. However, SQWP sees no conflict in his serving on the division level team. In fact, it is TVA's philosophy that safety is 52

, . , . , , r_, . , , ---.m-,-, , 7 --- % ,--. - - ,_m

line management responsibility. Consistent with that philosophy, since the FSG was involved in this incident, the FSG supervisor should be involved in the investigation. The division level accident report did i provide basic conclusions and recommendations in the area of industrial safety. The team concluded that this event did not meet the requirments of the '

agency's procedure and made that recommendation to NUC PR management.

The NSRS response in reference A.29 stated:

We still disagree with the position of having the responsible supervisor participate as a member of the investigation team. This situation can create a perception of potential conflict of interest in terms of having an individual investigating an incident in which he and his crew were involved. We concur fully that safety is line management responsibility; however, appropriate feedback can be received from the investigative process without participating as a member of an investigation team.

l The Director of Occupational Health and Safety in reference A.38 wrote to the Manager of Nuclear Power and stated:

In regard to the composition of investigative teams for serious accidents, our view remains unchanged from 4

that outlined in the SAIP, and we do not plan to recommend any modifications. The need for a degree of independence on the investigative team, apart from the immediate organization in which the event occurred, is recognized in the SAIP and is a commonly followed practice in the safety profession.

For this follow-up review, it was determined that the Nuclear Power i

procedure for Serious Incident Investigations was issued May 14, 1985, and incorporated into SQN site procedures SQS 29 Which was approved July 1985. The independence of the investigation team is adequately given in section 4.2 of SQS 29 which defines an Independent Investigation Team as:

A team designated by the Manager of NUC PR that has defined responsibilities for investigation and report preparation. Team members are normally selected from organizations which do not report administratively or functionally to management located at the affected site.

53

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.me..n-.-

! Section 5.1.7 of SQS 29 adequately addresses scene . .

preservation as:

j The director of the affected site shall ensure that the scene of the incident remains undisturbed until arrival of the-investigation team, except as necessary to protect people and property.

The performance and reporting of investigations in section 6.7 of SQS 29 define timeliness, root cause evaluations and reporting content as:

Once the investigation team is formed, an

investigation shall proceed immediately and a preliminary report containing only factual data shall '

be prepared and transmitted to the Manager of NUC PR normally within 5 working days of activation of the team. This report shall form the basis for a decision to continue the investigation.

An evaluation report shall be prepared by the team, normally within 15 working days after the team is activated. This report shall provide an indepth l analysis of the root causes and define needed corrective actions. This report shall be presented to the affected site management at a briefing. .After the briefing, site management will have an opportunity to comment on the evaluation report and provide proposed actions. A final evaluation report containing the i factual account of the event, the team evaluations and defined corrective actions, site management comments,

, and the proposed management actions shall be prepared l by the team and transmitted to the Manager of NUC PR and affected management. This will normally take place within 25 working days after the team is activated.

A briefing of the Manager of NUC PR by representative of the investigation team and affected site management will normally be held to discuss the final evaluation report. At the conclusion of the investigation, the Manager of NUC PR shall define responsibilities and schedules for corrective actions.

It was determined that the reconumendation of the Nuclear power report that consideration be given to leaving the inner door open was not implemented. Instead, AI-8, " Access to containment,"

section 2.6, was revised to require informing personnel working in containment of airlock doors being made intentionally inoperable.

The instruction states:

The upper and/or lower containment airlock doors shall

not be intentionally made inoperable (prevent l personnel egress) while personnel are inside containment. If the doors must be made inoperable with personnel inside, they will be instructed to use an alternate exit.

54

Sections 4.2, 5.1.7, and 6.7 of SQS 29 satisfactorily address the i

elimination of conflict of interest, timely establishment of the investigation, allow sufficient time to conduct the investigation, to be thorough and to provide a factual account of the event. This ites 3 1

is closed.

i GG. I-84-12-SQN-20, Needed Reemphasis on the TVA and SON Employee Expression of Concerns for Safety and Safety-First Policies In the original investigation (reference A.25), NSRS determined that the employees did not relate their increasing concerns for the safety of the job to upper management and an expression of concern for the adequacy of the design of the new tool support base was not followed up. WSRS recommended that it should be emphasized to all SQN

, employees that they are responsible for voicing their views concerning safety. Also to emphasize to all supervisors, engineers, and foremen that responsible concerns expressed to them must be evaluated. The TVA and SQN safety-first policy should be emphasized i i

to all SQN employees that nuclear safety is the number one SQN

) objective and that safety first means before schedule and befoce

production.

The Nuclear Power response in reference A.28 stated SQN had numerous mechanisms available to the employee to express their concerns. The response concluded that SQW would, through normal safety comsunications, reemphasize the rights and responsibilities of employees as described in SQN Standard Practice SQS 7 and General a Employee Training GET 1.2.

For this follow-up review, NSRS conducted a review of safety documentation. Documentation related to safety responsibility has

' been in existence for some time as expressed in the occupational Health and Safety Manual; Administrative Instruction AI-30, " Nuclear Plant Method of Operation"; SQA 46, " Employee Complaints Concerning Safety and Health;" and others. The SQN Standard Practice SQA 7,

" Hazard Control Plan," specifically designates each employee to accept responsibility for performing duties using safe and reasonable procedures. It was determined that GET 1.2 no longer exists; however, GET 2.1, 2.3, and 2.4 primarily emphasize radiological safety and includes a presentation of the recently implemented 1

I Employee Concern Program (ECP) contained in Standard Practice SQA 178, "TVA Office of Nuclear Power Employee Concern Program Line

! Organization Procedure."

l The ECP, SQA 178, specifically identifies the Office of Nuclear Power i policy that safety and quality are paramount. In addition, the employee responsibility to identify safety concerns and the supervisor's responsibility to evaluate these is specifically l identified in SQA 178, section 4.0, which states:

All personnel involved in TVA nuclear activities have

' an obligation to protect the health and safety of the public and their fellow employees. To this end TVA I

has established the following policy regarding the i

l 55 i

E. ,_ _. _ _ _ -_ _ ._ _ ._ __ ._ .. _ . _

handling of information related to any condition, -

practice, or event which may adversely impact quality, deviate from technical or procedural requirements, or have the potential for degrading equipment, operating capabilities or personnel's ability to accomplish assigned responsibilities. Any such practice, condition, or event of which any TVA employee becomes aware should be brought to the attention of the employee's supervisor.

All supervisors have the additional responsibility for considering resolving, or referring such practices, conditions, er events brought to their attention.

I All Office of Nuclear Power personnel have received training on the Employee Concern Program. Other discussions on job safety are provided in sections IV. H. J. S. GG, and II (I-82-21-SQN-02, -05; I-84-21-SQN-06, -20, -22) of this report. This item is closed.

l 4

56 l

c

')Dt . I-84-12-SQN-21, Ineffective SON ISEG Activities In the original investigation (reference A.25), NSRS concluded that the SQW Independent Safety Engineering Group (ISEG) had been ineffective in performing the function that was originally intended

, for the organization. This was due in part to the dual responsibilities for compliance /ISEG activities and the lack of true independence from line responsibilities and pressures. The NSRS recommended that SQN reorganize or reassign functions as necessary to provide ISEG personnel adequate independence from line responsibilities and pressures. Idditionally, functions should be limited to ISEG-type duties as required by the Technical Specifications.

In reference A.28, Nuclear Power responded by stating SQN does not agree that a broadly stated conclusion can be justified based on the evaluation of a single event. The response also stated:

4 The SQNP ISEG organization has been described to NRC in correspondence and the site NRC residents are very I

aware of the ISEG organization. The prosent i organization is an effective means of meeting the intent of NUREG and technical specifications requirements. The line duties of the compliance staff (coordinating the plant's response to all inspection / audit findings, investigation of potential reportable occurrences (PROS), preparation of Licensee Event Reports (LERs), tracking of corrective actions, and trending of Pros, LERs, and NRC violations - in short the maintenance of a broad overview of all activities potentially impacting plant safety) serve

! to enhance not detract from the ISEG function. SQNP acknowledges that the ISEG was not directly involved in the discussions and preplanning associated with this specific maintenance activity. The size of the ISEG staff necessarily precludes its detailed involvement in the conduct of every maintenance and operational activity occurring at the plant. The

focus of the ISEC review activities in fulfilling its nuclear safety engineering function is directed toward determining the overall effectiveness of plant programs and systems Which affect nuclear safety. To accomplish this objective, the ISIG monitors trends and looks for possible generic deficiencies in plant programs and systems.

The Office of Nuclear Power has not identified any progransnatic problems associated with the SQNP ISEG function. This finding is supported by previous NRC, TVA Nuclear Safcty Review Board, and TVA Quality Assurance evaluations in this area.

57 s

4 r , - e - -. _ .. _ _ .. _ . , . _ _ . - . . . , . .

The WSRS response in reference A.25 stated: . .

The NSRS concurs with your observation that a broad conclusion regarding ISEG effectiveness should not have been drawn based on one event.

Your response does not directly addr'ess the question

}

of whether the ISEG, as structured, meets Technical Specification requirements of having at least five dedicated full-time. engineers onsite to perform this function. This issue will be examined in greater

detail by NSRS in a future review.

J The NRC Region II Inspection Report of the thimble tube event addressed the ISEG function in Section 11.0 of the report. With respect of the ISEG activities, the NRC concluded that:

The ISEG efforts appeared adequate, and their

technical decisions did not appear to suffer from being incorporated in the Regulatory Compliance J Group. While the five individuals comprising the ISEG i have a dual reporting requirement (onsite and j independent offsite), the inspectors found no obvious l lack of independence in the performance of their duties.

l The NRC report also addressed the ISEG staffing issue. It concluded that:

{

The inspectors and reviewers concluded that since the licensee appeared to be responsive to the need for STA training and since the ISEG appeared to be performing its desired function, that the overall intent of Technical Specification 6.2.3.2 was met. However, to avoid future questions, the inspectors indicated to licensee management that an alternate member should be i

assigned during excessive (one month) periods of absence. Licensee management agreed to implement this

, or similar guidance. No violations or deviations were identified.

i For this follow-up review, additional documention was reviewed and l discussions held with the SQu and Nuclear power staff. A proposed Technical Specification Change No. 111 was prepared and issued for review. Section 6.2.3.4 of the Technical Specification was proposed to be modified as a result of TVA organization changes to have the ISEG function report to the Site Director rather than the Assistant Director for Maintenance and Engineering of the Division of Nuclear Power (a position that was abolished by the reorganization). As a result of their review, the Nuclear Safety Review Board had the following concerns related to the proposed changes:

58

0 .

The board believes that these ISEG changes are substantive and should be specifically mentioned in the description of the proposed change which will be submitted the the NRC. Furthermore, the board believes that the proposed ISEG change does not meet NUREG-0737 in regard to (a) reporting offsite to a corporate official who holds a high-level technically oriented position that is not in the management chain for power production, and (b) composition of five dedicated full-time engineers. In this regard, the board believes that an adequate justification for the exceptions to NUREG-0737 should be provided for submittal to NRC.

Reference F.1 rubmitted the proposed Technical Specification change to NRC with: (a) the ISEG/ Compliance staff reporting to the Plant Manager with a dotted line ISEC function reporting to the Site Director, and (b) a change to Section 6.2.3.2 that deleted the word

" dedicated" i.e. to read "The ISEG shall be composed of at least five full time (dedicated *) engineers located onsite."

The SQN justified the management reporting change by providing the various responsibilities of the Site Director in that he is a high-level corporate manager, located onsite in a technically oriented position, and is responsible for all activities affecting the plant.

However, the Plant Manager remains directly responsible for day-to-day operations. The dual roles of compliance /ISEG were justified by stating that the tasks are complimentary and that a departure.from the total dedication to ISEG functions, specified in NUREG-0737, is justified. An additional change to the Technical Specification has been submitted in reference F.4. This proposed change shows the plant Compliance Staff (noting that the plant Compliance Staff fulfills the responsibiltiy of the ISEG) reporting to the Site Director.

• Word deleted from proposed change The plant Compliance /ISEG Staff currently has seven full-time engineers (six engineers plus one supervisor).

Six of the seven are STA qualified with the seventh scheduled to start STA training at the beginning of 1987.

The engineers rotate on STA shift work thus bringing the knowledge gained on plant operations to their ISEG function. This is considered to be an excellent practice and should be continued. It also satisfies the NUREG-0737 recommendation of integrating the STAS into the ISEC function to enhance the. group's knowledge of and contact .

with day-to-day plant operations. This staffing level is -

consistent with the five engineers specified in the ff Technical Specifications. ~,_ .

l -

e ' )[M . - -

, 59 -

- " yh..,2

.M/+**

' 7 " [ i 2 .*A Z{

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

I Four ISEG reports were generated from June 11, 1985, through January 30, 1986, and a memorandum thet plant operations management sent to the operating personnel i

incorporating several recosmondations of an ISEG report were reviewed. Based upon.the review and discussions with I' _

several of the ISEG/ Compliance engineers, it was concluded.

i that the ISEG reports demonstrated thst: (a) comprehensive and thorough reviews are being performed and documented, (b) root cause determinations and recommendations to prevent future occurrences are made, (c) the Plant Manager decides whether the recommendations will be implemented then they are put on the Corrective Action Tracking System (CATS), and (d) the evidence suggests that the reviews were

conducted in an independent manner.

S, Based upon the discussion with several of the i

ISEG/ Compliance Staff engineers and the review of the ISEG reports and the resulting reconumendations, it was concluded l that: (a) the independent performance of the ISEC function did not appear to be adversely influenced by reporting to a

the Site Director, (b) the dual ISEG/ Compliance Staff has at least five full-time engineers onsite during working hours; however, they do not devote 100 percent of their time to the ISEC function, and (c) the NRC has been kept

' informed of the current ISEG/ Compliance reporting and '

staffing arrangement by the proposed Technical

Specification changes submitted in October 1984 and November 1985 and by discussions with WRC. Also, changes i i

in the ISEG reporting chain are being strongly considered by the new Office of Nuclear Power top-level management.

The final resolution of this recommendation will be made-when NRC and TVA resolve the proposed Technical

' Specification combined with top-level Office of Nuclear - <

Power organisation changes that may impact the Office of Nuclear Power Manager to Whom ISEG reports. Since positive

] action has been taken by TVA which will be pursued to a

resolution with NRC, this item is closed.

1

II. I-84-12-SQN-22, Sinnificant Breakdown in the SON Procedure l Process for Maistenance Activities e

In the orldnal i westigation (reference A.25), NSRS >

! reconumended 'hmt-4 The procedural process for maintenance activities at SQN should be thoroughly evaluated. Corrective j actions including procedure verification should be

initiated as necessary to improve the (1) knowledge of i j

those personnel preparing and using procedures of What i

constitutes an appropriate procedure, the quality

' elements that should be incorporated into a procedure, and the change process for existing procedures; (2) quality of the PORC and biennial reviews; and (3)

] compliance with procedures.

l 60 i

. . _ _ _ . . . . . ~ , . - . - . . - . - _ - , _ . , _ . _ _ . , , . , , . , ....--,_..m.,_ m, . , . _ . ,, ,- , . , , , _ _,%.-- .-

The Nuclear Power response in reference A.28 addressed -

i recommendations I-84-12-SQN-05, -7..-11. -17, and -22 as one item.

The following is an extract that provides a description of their decisionmaking logic and procedural guidance:

. Management made the decision to clean the blocked thimbles tubes while at the 30 percent power level and specified adequate guidelines and precautions to conduct this work activity. However, the work package (MR and Special Maintenance Instruction SMI-0-94-1) were not revised to reflect those directions.

Discussions were held between the cognizant engineer and foreman concerning the high pressure connections and their proximity to the 10-path breakdown connections. No work was to be done nor was it done without the lead engineers at the seal table. The 10-path transfer devices were disconnected and rolled back prior to beginning the cleaning process without an MR or procedural guidance, but the engineers involved were aware of the unit conditions at the time of the work, the system design, mechanical makeup of the components, and potential hazards. Employee i

awareness of the unit conditions and absolute

requirements was demonstrated by informal planning and

< cursory attempts at satisfying requirements. The

! at-power cleaning process began using the MR and l

SMI-0-94-1 as procedural guidance.

i i For this follow-up review, additional information was obtained by

~

having discussions with SQN personnel, reviewing draft procedural

guides, and the NRC Syctematic Assessment of Licensee Performance j (SALP) report for the period March 1, 1984, through May 31, 1985.

The SALp report provided the NRC assessment of the overall maintenance process. The following is an extract from the report

l The overall quality of maintenance operations has been

! erratic, ranging from poor to good. .. . Some technicians performing maintenance tasks were observed using good work practices and implementing the

, management expressed philosophy of adhering to l procedural requirements; however, maintenance insttvetions were weak or nonexistent for some-safety-related activities. Several procedural violations were identified during the assessment period . . . involving the failure to establish or I implement procedures. In general maintenance l

procedures were adequately written and followed.

However, many procedures were cascaded and interwoven with other procedures, requiring technicians to transfer between documents in order to complete a single maintenance' activity. There was also some duplication of procedures written for similar tasks by different organizations (Office of Nuclear Power, OE). The cumbersome procedural interdependencies resulted in confusion.

61 l

1 4

Discussions with supervisors and engineers indicated that this was a l 1

fair assessment for that time period. The review of the NRC report on the inspection conducted from December. 2-6, 1985, indicate that

significant improvements have been made at SQN, since a team of 11

inspectors identified only 3 examples of failure to follow procedures in the areas of motor operated valve modification.

{ As a result of the SALp report, meetings drith foreman and craft were '

j held during November 1985 to inform them on how to use plant instruction change (ICF) forms in order to have existing procedures l changed when errors were found or the instructions were inadequate to i- parform the activity. Several recently revised mechanical j maintenance instructions were examined. Most had been instituted by j ICFs originated by mechanical maintenance engineers, but one dated 1 December 12, 1985, was originated by a mechanical maintenance l planner. This indicates that instruction details are being reviewed I

and the ICF system is being used.

Knowledge of personnel preparing procedures was verifled by reviewing maintenance instruction in the draft stage. Two  ;

mechanical engineers, both degreed engineers, had written MI 10.05.1 and MI 1.11. Both procedures appeared to be satisfactory for the ,

j activity to be performed and were being sent for review. One had j

used the draft SQN writers guide and had found it to be a useful I

tool. It was noted that the craft are now required to review the draft instructions in addition to those previously required to do so. The results of instruction reviews by the craft were examined.

, It was concluded that the craft suggestions for improving the

)

instruction were positive and practical.

a The biennial review of Maintenance Instructions (mis) and Instrusent j Maintenance Instructions (IMIs) was examined. The newly appointed 3 Mechanical Maintenance Supervisor had found that the comunencement of 1 the review of mis was overdue and had issued a deviation report

}. DR# 85-10-127 R. A three-page checklist was developed for the MI.

review which was completed on December 30, 1985. Changes to the

existing mis have been postponed and proposals from external contractors to rewrite mis are currently being studied with no decisions having been made. In addition, a consnitment has been made i

to NRC for SQN to review all mis with a fully developed checklist by July 1987.

1 i A draft procedure writers guide has been developed by mechanical i maintenance using INp0 85-026 and NUREG/CR-139 as guidelines. It is 4 proposed to become Appendix A of SQM 1. The Instrument Maintenance i

Section has developed a checklist for review of plant precedures and issued it as IMS-I32 on November 27, 1985. The quality Assurance i Section Instruction Letter SIL 5.1 in effect at the time of the

{ thimble tube event was compared with that currently used. A procedures review checklist has been added since that time. That checklist has also been. incorporated into IMS-I32.- These checklists l - and writers guides will improve the preparation and review quality of

! procedures.

i 62 i

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~

Discussions were held with the QA PORC representative and the.QA .

representative on the PORC subcommittee for maintenance procedure reviews. It was their opinion that both the quality of proposed procedures and procedure reviews has improved. The quality of PORC reviews was assessed by examining current PORC reviews of maintenance instructions. The PORC subcommittee for the teview of IMis in accordance with AI-4 had met January 31, 1986. The subcommittee recommended in their memorandum dated February 3, 1986, that their report be accepted. The report was reviewed and rejected by PORC.

At the time of this review, the report had been revised (but not submitted) to incorporate the PORC comments. This indicates that PORC is doing the job it was intended to do.

Compliance with procedures has been and is being stressed on a continual basis to SQN personnel. SQA 129 was issued by the plant manager in January 1986, and it stresses compliance with instructions and to take time to correct those that are inadequate. In addition, at the Plant Managers' daily morning meeting, failures to follow procedures are discussed. Additional details are provided in J sections IV.R and IV.GC (I-84-12-SQN-05 -20) of this report.

The implementation of checklicts for procedure reviews, involvement of all levels of personnel in the reviews including the crafts, commitments have been made to NRC to r3 view all mis by July 1987, i foreman and crafts were trained on use of plant instruction change forms, tangible evidence exista that PORC reviews have improved, and compliance with instructions is emph. ized on a continuing basis satisfactorily resolve this recommendation. This item is closed.

4 JJ. I-84-12-SQN-23. Inadequate Reporting of the Event to NRC In the original investigation (reference A.25), NSRS recommended that SQN revise the LER to reflect the true nature of the leak, the adequacy and violation of SMI-0-94-2, and the effective long term corrective action.

The Nuclear Power response in reference A.28 was the following:

The true nature of the leak (cate, amount, duration, its effect on instrumentation, as well as the ferrule l failure and thimble ejection) was adequately described.

The LER did not mention inadequate procedures or

! failure to adhere to procedures in conduct of the i

maintenance activity because the plant did not and does not consider these to be causal factors of the

event.

! The LER will be revised by submittal of supplemental information to the NRC to indicate the cleaning

! technique in use at the time of the event will not be performed with the reactor coolant system at temperature and/or pressure but that other available 63

] . -

techniques will be esrefully and thoroughly evaluated i

prior to any future decision to clean thimble tubes t

with the reactor coolant system at temperature and/or pressure. In addition, the Office of Nuclear Power response to this NSRS report will be included in the supplemental LER submittel to NRC so that the full scope of short-term and long-term' corrective actions associated with all aspects of this event are brought to the attention of NRC.

The NSRS response in reference A.29 concluded that the LER was not j complete because of the following:

The manner in which the leak initiated and the rapidity with which it escalated was not accurately described. This description is necessary for a

complete understanding of the event [see 50.73 (b) (2)

(i)).

l Regarding lack of adherence to the special maintenance instruction SMI-094-1. NUREG-1022, on pages 18 and 26, and Question and Answer 2.7 on page 5 of WUREG-1022,

, Supplement 1 specify that violations of procedures are to be I

} reported in the LER regardless of whether such violations are causal factors.

The ESRS response further stated that:

The NSRS recommendation on this item was to revise the LER. Inasmuch as you have connaitted to revise the LER and to submit the Office of Nuclear Power response and the NSRS report to the NRC, we believe these il submittals will meet the intent of the recommendation.

i For this follow-up review, it was determined that the LER was revised and submitted to the NRC in reference F.2. The revised LER: (a) modified the original information on'the specifies of the I

instrumentation failures and calibration shifts; the class IE 1 qualified instruments experienced calibration shifts, one of'Which. l was outside the technical specification limit, (b)~ stated that'the i

modification of the orginal Teleflex tool (base added) was the basis of error for the event, (c) two long-term corrective actions were j' identified, and,(d) the NSRS report I-84-12-SQN and Nuclear Power response (references A.25 and A.28, respectively) were attached to i

the LER, thus both documents became part of the LER and public record. The specifics of the LER were not modified to revise the i description of the nature of the lead nor the adequacy 'and violation of SMI-0-94-1; however, since the revised LER attached the NSRS .

report and the ONP response it resolved the intent of these items. I l

64 I l

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4

. o The NRC report of the investigation of the thimble tube event, ,

concluded that the inadequacy and failure to follow procedures did not appear to be the true cause of the event and thus reporting requirements were not violated. The NRC report of the investigation of the thimble tube event, reference E.1, stated:

i l The inspectors reviewed 10CFR50.73, the 10CFR50.73 l

statements of consideration, EUREG 1022, and handouts

! from NRC-sponsored seminars on the new rule. The inspectors determined, particularly from the

' statements of consideration, that licensees should report personnel errors and inadequate procedures associated with reportable events when those errors or procedures caused the event or impeded the recovery from the event. Based on the preceeding paragraph of

!, this inspection report, the true cause of the event appears to be use of an improperly modified tool.

While the inadequate SMI and failure to follow or 4

change the SMI show inadequate understanding and implementation of NRC procedure establishment and procedure compliance requirements, it appeared that, had these deficiencies been corrected prior to conuencing work on the thimble tubes, the tool would likely still have caused the leak event.

Consultations between the inspectors and AEOD I supervision also supported the conclusion that

reporting requirements were not violated by the licensee's decision that these deficiencies were not i

pertinent to the event.

) It is concluded that corrective action has been taken to resolve the recosmondatior. since the NRC has determined that the SQN LIR did not violate the NRC LER reporting requirements (it is the judgment of NSRS that the procedure

, violation and inadequacy factored into the event initiation to somes degree), SQN has deleted the instruction in question, procedure compliance has been stressed to all personnel, and a revised LER was submitted to NRC with the 4

NSRS report and Nuclear power response as an attachment.

This item is closed.

! KK. R-84-17-EpS-02, Lack of Approval of onsite Vendor Services at SON In the original review (reference A.41), NSRS recomunended '

i that SQN should develop and implement a program that satisfies the requirement and intent of 0QAM [NQAM), part III, section 2.1, paragraph 10. For the follow-up review, SQN provided additional information'which was reviewed.

} The NSRS report cited three examples of vendor service for which nc QA documentation was provided by the site (though repeatedly requested by NSRS) that demonstrated the work was accomplished in accordance with the QA requirements.

} 65 b __ _ . _ . _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _

• . \

.g Because the documentation was not provided at the time of the review, NSRS had to assume none existed; therefore, the WQAM requirement was not being met. '

In the May 21, 1985 (L12 850520 800) Nuclear Power response to the report, some documentation was provided to demonstrate that proper QA was provided on each of the three examples. That response, however, was insufficient.

In the September 5,1985 response (L12 850826 800),

additional information was provided. Based upon that information, a review of valve drawings D-3-1500-11 R8, 47W809-1 R26, 47A366-62-11 R9, and telephone conversations with the vendor, Crosby Valve and Case Company, on one of the examples, it was determined that proper quality control was applied to the vendors and adequately monitored by SQN. This item is closed.

LL. R-85-02-SQN/WBN-01 (NUC PR) Office-Wide Awareness Bulletin for Tube Fittinz Maint nance Activities In the original review (reference A.42), NSRS reconunended that a NUC PR office-wide awareness bulletin or similar mechanism should be prepared and distributed to the nuclear plants. The bulletin should discuss tube fitting design; assembly, reassembly, and inspection criteria; policy on interchanging components; failure modes (including those identified by the SQN and W8N maintenance craft personnel);

hazards involved in working on pressurized fittings; and should specify s'pecial precautionary measures when maintenance on pressurized fittings is necessary. The desired bulletin should be incorporated into a permanent instruction at each plant for future awareness of new employees. ,

i The Nuclear power response in reference A.44 stated that:

We are in the process of preparing an office-wide awareness bulletin to address tube fitting design;

' assembly, reassembly, and inspection criteria; policy on interchanging components; failure modes (including those identified by the SQN and WBN maintenance craft personnel); hazards involved in working on pressurized fittings; and to specify special precautionary measures when maintenance on pressurized fittings is necessary. The bulletin will be distributed to all plants for incorporation in plant instructions and -

training.

For this follow-up review, the~ awareness bulletin and training documentation was reviewed in addition to having discussions with .

POTC SQW. WBN, BFN, and BLN personnel to determine the status of craft training programs identified in the awareness bulletin.

66

.9

_ _ _ _ _ _ _ _ _ _ _ - . - _ _ _ _ _ _ - - - - - - - - - - - - - - - . - - - - - - - - - - - --- i- - -

e The office-wide awareness bulletin was sent to BFN, SQW, and WBN management; however, it was not sent to BLN. During this review, telephone conversations were held with the BLN Maintenance Superintendent and Mechanical Maintenance General Foreman and it was determined that the office-wide awareness bulletin had been recently received by them. It was also determined that SQN had developed Hazard Control Instruction HCI-M23. " Tube Fittings," which is as an integral part the Nuclear Power bulletin. The bulletin references NRC IE Information Notice 84-55 which describes the significant events at Zion Generating Station and the SQN thimble tube ejection.

The Zion event occurred when a fitting (SWAGELOK) " broke loose" at the guide tube causing an unisolatable lesk of reactor coolant. The cause of the event was attributed to the fitting ferrule assemblies in most of the guide tube being displaced from their original position. The NRC notice indicated that in both events maintenance was being conducted on a high-pressure system with what was equivalent to single valve protection (the fitting). The bulletin also states that hazards associated with pressurized fittings (high temperature and high pressure) are compounded in a nuclear plant by the possibility of contamination. These two events described in the bulletin emphasize the importance of proper tube fittings and assembly.

The awareness bulletin also provides a policy statement that compression-type tube fittings shall be installed consistent with the manufacturer's instructions; maintenance activities involving these fittings shall not degrade the integrity of these fittings; fitting components made by different manufacturers or tube fitting components which are different types made by the same manufacturer are not to be interchanged with each other. The awareness bulletin provided a brief description of the POTC training class on proper tube fitting and provided a brief summary of the class. The training class summary as described in the bulletin is as follows:

1. Discusses SQN practice which states, " tube fitting components made by different manufacturers or tube fitting components which are different types but made by the same manufacturer are n_ot interchanged with each other."

2. Emphasizes the designs of different brands of fittings and the proper orientation of components.

3. Provides specific preparation instructions for proper assembly, disassembly, and subsequent reassembly of pressurized tube fittings (i.e., tube cutting, bottoming tube in fitting body, tightening).

l 4. Discucses inspection and use of "SWAGELOK" gap inspection gauge.

In summary, the awareness bulletin addresses the following aspects of the recommendation: (a) identifies hazards, (b) provides the policy on interchanging components, (c) and special precautionary measures when maintenance on pressurized fittings is necessary. The training 67

l program also covers precautionary measures. Additional precautionary

• measures are described in the thimble tube maintenance instniction

• MI-1.9, "Botton Mounted Instrument Thimble Tube Retraction and Reinsertion," and MI-1.10. "Incore Flux Thimble Claaning and Lubrication." These precautions are identified in section 3.0 of MI-1.9 and section 4.0 of MI-1.10. The specific one related to the hazards of pressure and temperature is covered by the precaution (in both procedures) that:

There is to be no maintenance on the high pressure fittings while the primary system is pressurized above atmospheric or head pressure from inside the guide tube. If there is to be any tightening or loosening of the fittings with the primary system above atmospheric or head pressure a unique procedure reviewed by PORC and approved by the plant manager is required.

In addition, procedure MI-1.9 prerequisite 2.1 requires that:

All personnel working on tube fittings should have had the tube fitting class.

Other instruction issues related to SWAGELOK-type fittings are discussed in detail in section IV.191 (I-85-02-SQN/WBN-02) of this report. In summary, the thimble tube instructions have been modified (subject to recommendations in section IV.let of this report) to cover the precautionary measures on maintenance and cleaning of the thimble tubes, the hazards and risks involved in working on the pressurized fittings have been removed by not performing cleaning and lubrication or thimble tube retraction and reinsertion if the system is above atmospheric pressure.

The SQN tube fitting training was identified in a July 22, 1985 memorandum from B. M. Patterson to Robert H. Harris, " Tube Fitting Class," that stated the SQN plant had decided to require the tube fitting class as a qualification for steamfitters who will be responsible for initial installation of tubing systems and for selected cognizant engineers and inspectors. The course material was developed by POTC and SQN personnel and is documented in manual letT-28 " Student Manual, Initial Tube Fitting." The training material and awareness bulletin were transmitted to the BFN, WBN, SQN, and BLN sites. .The BFN maintenance training supervisor stated that it was planned to train the instrument mechanics, moohanical steamfitters, and machinists at:the rate of two sessions per day (availability of personnel permitting) until the required personnel had completed training., SQN training has been going on for a long period, and a significant number of instrument mechanics, quality assurance, modifications, mechanical maintenance, engineering, and test group personnel have completed training. The BLN training had not been-initiated, but a program was being actively developed with POTC. Based upon discussions with the WBN Maintenance Training Supervisor, training was being conducted for maintenance, modifications, construction, and Nuclear Service Branch personnel, j and was about 90 percent complete.

68 l

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

The letT-28 training course incorporates the NSRS recommendations in ,

the following manner:

1. Tube Fittina desian - Segment II, " Fitting Identification and

' Installation," has photographs of parts in disassembled / assembled configurations for SWAGgLOK, Parker CPI

!- Type BZ (Short Parker), Gyrolok, Parker Forulok Type BU (Long Parker). Imperial-Eastman (Hi-Seal), and Tylok fittings.

] 2. Assembly-reassembly: Segment II, Sections II IV, V. VI, VII,

and VIII provide instructions on disassembly-reassembly of the i

fittings. In addition to the instruction material practical 4

training was conducted for the craftsmen on the course work.

Retightening procedures are provided, -as stated in Appendix B.

3. Inspection Criteria - Segment II,Section III, specifically j' addresses the SWAGELOK gap inspection gauge and describes its use, l

4. Policy on Interch===ina Cenents - page 1 of the introduction specifically states:

It is the policy of P&E (Nuclear) that tube fitting cosponents made by different manufacturers, gg tube fitting components made by

the same manufacturer which are different, ARE I NOT INTERCHANGED with each other. Use the same

' brand and type nut, ferrule, and fitting body for .

each individual tube connection. . . . This does mean that all individual components of a single tube connection will be the same brand and tree.

For example, if you are using a short Parker fitting body, use a short Parker nut and ferrule.

5. Failure Modes and Hereeds - This is addressed by the statement:

j Improperly installed tube fittings may seal well enough to hold a limited amount of pressure sometimes to the point of passing a hydro-test; i

however, the installation will not be

• i 1

mechanically sound and will not withstand the  : '

_ vibration that occurs once the system is

operational.

'. Also, procedure revisions on thimble tube to retract and clean at atomspheric pressure precludes temperature pressure hazards. At the 1

training session the instructor provides a description of some j significant safety events that have occurred in the industry with tube fittings to depict the types of problems that are occurring. At i

the end of the training session, practical and written exams are given. A 70 percent grade is considered passing.

69 l

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l .

. . 1 With respect to the recommendation that a permanent instruction at ,

each plant for future awareness of new employees be implemented, the SQN maintenance practice require the use of craftsmen trained in IRIT-28 for work conducted on tube fittings. The SQN IMPO Accreditation Self Evaluation Report dated January 1986 for Mechanical Maintenance Craftsmen Training also addresses this issue.

In the mechanical maintenance area, the responsibility for ensuring that only individuals do are qualified to perform tasks independently is delegated to the foreman and general foreman by i sections 5.3 and 5.4 of IEESL-A65, " Mechanical craft Training Program." The supervisors and foreman maintain a current listing of training qualifications for the craftsmen to ensure that qualified individuals will perform the work. The " Task-to-Training Matrix" of the INPO Accreditation Safety Evaluation Plan for Craft Training, ,

Tasks SFP 112. " Install Steel Tube Fittings," and SFP 113. " Install Copper Tube Fitting," identifies letT-28 training as a requirement to perform this work. Based upon the training requirements to perform j

tasks independently, new employees will not perform tube fitting on safety related components without appropriate training.

It is concluded that the awareness bulletin, procedures, and the INp0 accreditation craftsman training program satisfactorily incorporate all of the aspects of this recommendation. This item is closed.

l 131. R-85-02-SQN/WBN-02, Maintenance. Operatinz and Test Instructions j In the original review (reference A.41), NSRS reached the following

conclusion concerning the SQN incore instrument tubing seals.

l Instructions at SQN did not contain sufficient i clarity, precautions, warnings, and other measures to provide the desired level of confidence that the high-pressure mechanical seals will not be degraded during maintenance. activities or to lessen the severity of the consequences of a failed seal.

NSRS made the following recommendations:

Applicable maintenance, operating, and test instructions should be revised as necessary to provide j consistent guidance for system assembly, reassembly, and inspection of all SWAGELOK and mixed fittings; address replacement of ferrule assemblies on previously undisturbed tubing; address lubrication and inspection of fitting threads to minimize or detect wearing, galling, and cross-threading; specify  :

limiting forces d ile using the low-pressure seal; add -

cautions and warnings against interchanging fitting a components, cross-threading, turning of fitting- _

bodies, excessive forces, working on seals d ile the primary system is pressurized above atmospheric, and 1 increasing primary system pressure d ile thimble tubes i are disconnected from the overhead. path transfer system.

70 ,

I

i For this follow-up review, NSRS reviewed the applicable maintenance, operating and test instructions, and a workplan, and interviewed mechanical maintenance personnel. The individual reconumendations were addressed as follows:

Consistent muidance for system assembly. reassembly. and inspection

of "*'MK and mixed fittinas: The necessity for separate criteria for mixed fittings and fittings other than SWAGELOK has been removed by ensuring that all the fittings are SWAGELOK. The original installation on unit 2 was all SWAGELOK and the changeover to SWAGELOK on unit I was completed in late 1985 by workplan 11818 (ECM 6537). Maintenance Instruction MI-1.9, " Bottom Mounted Instrument Thimble Tube Retraction Reinsertion," revision 7, and special maintenance instruction SMI-0-94-3, " Seal Table High pressure

, Seal Repair," revision 1, both include the appropriate instructions j and criteria for assembly, reassembly and inspection of SWAGELOK i fittings. Special maintenance instruction SMI-1-94-5, " Thimble Tube i Installation," revision 1, is not intended to be used to remake the i high pressure semis, but it does not state that this is the case. It

! includes the SWAGELOK inspection gauges in the list of tools and work aids, which implies that the fittings are to be remade as part of the activity covered by this procedure. A proposed new instruction,

! MI-1.11. " Thimble Tube Installation," when issued will replace SMI-1-94-5 and should fix this probPua. The draft of MI-1.11 reviewed includes the appropriate instructions for making up the high pressure seals.

I Replacement of ferrule assemblies on previously undisturbed tubint:

MI-1.9 and SMI-0-94-3 both require installation of new ferrules on a previously undisturbed surface of the guide tube.

Lubrication and inspection of fittina threads to minimize or detect wearina. mallina. and cross-th Impding: MI-1.9 requires inspection of the threads of high pressure fittings for signs of galling, wearing, or cross-threading, and application of NEOLUBE to the threads.

SMI-0-94-3 requires inspection of nuts and reducer union bodies and ,

replacement if damaged, but does not specify inspection for galling, wearing, or cross-threading. This is acceptable because visually detectable galling, wearing, and cross-threading are " damage," and because this instruction would normally be used as a result of

, activities conducted under MI-1.9 which specifies the inspections.

i SMI-0-94-3 does not specify the use of a thread lubricant, but states

under precautions that a coat of NEOLUBE may be applied to the-threads. NSRS believes that a thread lubricant should be i consistently used. A proposed revision to SMI-0-94-3 requires the use of thread lubricant.

Specify limiting forces while usina the low pressitre seal: MI-1.9 specifies the method of tightening the low pressure seal to achieve proper torque and cautions against overtightening due to possible s

ferrule damage. This is the only procedure that addresses use of the l low pressure seals, i

71 t

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. ~

Cautions and warnings amainst interchanninn fittina components: All fittings are now SWAGELOK and applicable procedures use the name SWAGELOK frequently so that it should be obvious that other fittings are inappropriate. Training provided to the crafts provides policy on not interchanging components. This is also discussed in detail in

! section IV.LL (R-85-02-SQW/WBN-01) of this report.

Cautions and warninas amainst cross-threading: Both M.I-1.9 and j

SMI-0-94-3 address assembly of compression fittings, but neither i caation against cross-threading. Cross-threading potential and I checking for cross-threading is covered in detail in the craft training program MNT-28. This is discussed in section IV.LL l (R-85-02-SQN/WBN-01) of this report.

Cautions and warnings against turninz of fittinz bodies: MI-1.9 i

includes the appropriate caution when disconnecting and remaking the fittings. SMI-0-94-3 does not presently caution against allowing the fitting body to turn, but the proposed revision to this procedure does include this caution. The proper disassembly and reassembly procedure for fittings is also provided in the crafts training program MNT-28. This is discussed in section IV.LL (R-85-02-SQN/WBN-01) of this report.

Cautions and warnings amainst excessive forces. - The intent of this recommendation was to ensure that excessive forces both apparent such as over-tightening, and less obvious, such as bending due to improper use of a wrench would be considered. MI-1.9, MI-1.10, SMI-0-94-3, and possibly special maintenance instruction SMI-1-94-5 should include such precautions. Presently, they do not include any precautions against excessive forces except for the cautious against overtightenir.g in MI-1.9 and SMI-0-94-3. At the exit meeting the plant agreed to review the thimble tube maintenance instructions for steps that could apply excessive forces. This review was documented in a memorandum (referenes A.78) which states:

. . (MN outage and major Maintenance Support supervisor) and . . . (Mechanical Engineering) have review the tasks involved in the following MI's/SMI's:

MI - 1.9 MI - 1.10 MI - 1.11 (Replaces SMI l-94-5)

SMI 94-3 In their expert opinion, all steps which could apply excessive forces to high pressure seals have sufficiently detailed and clear instruction to control the possibility of over streesing or compromising the I integrity of the high pressure seals.

i Proper assembly techniques are also covered in the craft training program MNT-28. This is discussed in section IV.LL (R-85-02-SQN/WBN-01) of this report.

72 l

~+ ..

i Cautions and warmin== amainst workina on the hiah oressure seals while the primary system is pressurized above atmospheric: MI-1.9

-and MI-1.10. "Incore Flux Thimble Cleaning and Lubrication," revision 3, include appropriate precautions. MI-1.9, MI-1.10, and SMI-0-94-3 include prerequisites that the reactor be in mode 5 or 6. SMI-0-94-3 should include the same precaution that appears in MI-1.9 and MI-1.10.

cautions and warnints maainst increasina crimary system pressure while thimble tubes are disconnected from the overhead path transfer system: . None of the applicable instructions include this precaution. ESRS considers this extremely important because it would prevent complete ejection of a thimble tube in the event of a high pressure seal failure. Inclusion of this precaution in GOI-1, " plant Startup from Cold Shutdown to Hot Standby," would be adequate, but other methods of addressing the problem mayalso be appropriate.

This item remains open pending completion of the following items:

1. Issuance of the proposed MI-1.11. " Thimble Tube Installation,"

which will repisce SMI-1-94-5, and addresses several of the original reconumendations.

2. Issuance of the proposed revision to SMI-0-94-3 that to require the use of an appropriate thread lubricant, and cautions against allowing fitting bodies to turn.

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

4. Revision of appropriate instructions to preclude pressurizing the primary system withthe thimble tubes disconnected from the overhead path transfer system or at least preclude any work on the seals with the primary system pressurized above atmospheric knd the thimble tubes disconnected from'the overhead path transfer system.

The above procedure revisions should be made prior to the next use of the procedure.

NN. R-85-03-NPS-01, Inadequate Definition of Responsibility In the original report (reference A.54), MSRS concluded that the responsibility for determining the identification and availability of spare parts was not clearly defined in SQN. procedures. The

^

recommendation was to procedprally define this responsibility." The Nuclear power response said'that:'

SQN has adequately defined the responsibility for identification and availability of spare parts. The job description for the maintenance planners is the most definitive document but is not comprehensive due to the diversity of problems associated with

.73

. o material. Planners, foremen, craf tsmen, and engineers - .

all have some responsibility for the various aspects of identification, availability, and location of materials depending on the complexity of material / parts specification, priority of the job, and whether the items are being obtained from Power Stores or being procured from a vendor.

ISRS commented on the response by saying that if the plant chooses to define this responsibility in job descriptions (such as in the draft for maintenance planners included in the response) rather than in a separate document, this is acceptable.

For this follow"up review, current job descriptions for four maintenance planners (two mechanical, one electrical, and one instrumentation) were reviewed. The four job descriptions were identical and did pisce responsibility for identifying and determining the availability of spare parts / material with the planner. One of the mechanical planners and the electrical and instrumentation planners were interviewed concerning their responsibility in this area. They were all aware of their responsibility for identifying and determining the availability of spare parts. It should be noted that generally the identification of spare parts does not occur in the planning phase because most of the maintenance are for repairs rather than scheduled maintenance. This item is closed.

00. R-85-03-NPS-04, ASME.Section II Postmaintenance Valve Testinz - SON In the original review (reference A.54), NSRS determined that the SQN Instrument Maintenance Section did not identify the need for ASME Section XI valve testing when o they performed work on Section II valves and recomunended "

they be trained in the purpose of the ASME Section II pump and valve program and in how to identify pumps and valves which are included in it.

The NUC PR response was:

Our investigation has determined the three MRs identified in Section III.C.3.a of the subject report appear to be isolated cases for Section II postmaintenance testing. The valves listed,

! 1-PCV-01-12 and PCV-01-5, are tested using SI-166.3, '

"rull Stroking of Category 'A' and 'B' Valves During

. Cold shutdown," and the unit must be in mode 5 in l order to perform SI-166.3 for these valves. For the examples cited, unit I was in mode 5 only when Mas A-245631 and A-288723 were worked. However, the 74

• e Instrument Maintenance Section has emphasized to l

, section personnel the importance of identifying l postmaintenance testing requirements on MRs.

i i NSRS commented on the response as follows:

i The corrective action'taken by the plant appears to be adequate. However, the mode infor:aation included in

! the response is misleading. The fact that the unit 1 may not be in the mode necessary to perform the I Section XI postmaintenance testing does not waive the need to schedule the required testing when the plant

is in a mode that would permit testing.

For this follow-up review, NSRS interviewed the instrument j maintenance planners and deternined that they were well acquainted j with the ASME Section XI program-for valves as it affects the Instrument Maintenance Section. There are no pumps in the Section XI program for which instrument maintenance is responsible. The i

planners were familiar with all the pertinent documents including

SQM 2. " Maintenance Management System," TI-69, " Summary of pre- and j post-Maintenance Valve Tests for ASME Section XI and 10CFR50 Appendix j J " and the SI-1.66 series. They were using an informal checklist j they had prepared to assure that they considered all the necessary i items when planning work. All the planners understood that Section j II postmaintenance tests must be specified even if the plant is in an j operating mode that precludes the necessary testing at the time the 3 maintenance is done. In addition to the Section XI awareness of the planners, the instrument mechanics are made aware of Section II

requirements by training. This training is provided as part of an

annual' training course on pneumatic valve stroking. NSRS reviewed 1

1 the training course and found it to be appropriate and very comprehensive in its discussion of Section II requirements as they i

' relate to the work of the Instrument Maintenance Section. This item is closed.

] PP. R-85-03-NPS-06, Postmaintenance Testinz Program-Generic i

j In the original review (reference A.54),'NSRS determined that there l were no guidelines to ensure that postmaintenance testing instructions verify that the component or system worked or still

~

j j functioned as designed. NSRS recosmonded that each site prepare an j instruction outlining the criteria to be followed in selecting or

preparing postmaintenance tests.

l The response from SQN was:

l j PMT requirements are covered in surveillance

instructions and not by the Special Test Program as i indicated in the report and reconenendation section.

l Maintenance requests are reviewed by appropriate plant 1 personnel prior to commencing work to determine PIff requirements. The appropriate surveillance testing, 15 l

I f

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

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

f as identified in the technical specifications, aust be , ,

completed prior to declaring the component or system t operable and returning tc service.

i In commenting on this response, NSRS noted that it did not address

the concern expressed in the recosunendation. However, NSRS noted that the SQW response to an NRC level IV violation (50-327-328/85-24) may address the problem. This response was

A Stendard Practice SQM-2, " Maintenance Management System," will be revised to include direction for when PMT is required for maintenance activities and descriptions pl 3that should ha reauired in the P!IT. >

Personnel who plan maintenance requests will be instructed in proper PMT requirements.

f i

For this follow-up review, NSRS detemined that Change No. 85-1699,

' issued December 31, 1985, does add appropriate criteria for selecting

and preparing postmaintenance tests to SQM 2, " Maintenance Management i System." All the maintenance planners were trained to the procedure l

change. Several maintenance planners were interviewed concerning postmaintenance testing and all were aware of the need to verify proper functioning of the component or system af ter maintenance, and 4

all were familiar with the criteria in the change to SQM 2. Ten

recent maintenance requests were reviewed and all were found to have j had adequate postmaintenance tests specified and required. This item is closed foc SQW.

j QQ. R-85-03-NPS-07. Common Mode Failure-Generic 4

In the original review (reference A.54), NSRS determined that the l Mechanical Maintenance Sect. ion had no method of avoiding common mode failure. NSRS recommended that a program be developed and ,

impleannted which provides a method of avoiding cosmon mode failure.

Nuclear Power responded as follows:

SQW will implement a program in accordance with WQAM, Part III, Section 7.3, " Common Mode Failures,

j. Maintenance-Initiated." The program will be t implemented by January 1, 1986.

l For this follow-up review, NSRS determined that Mechanical l Maintenance Section Instruction Letter INISL-A36, "Cosumon-Mode Failures, Maintenance-Initiated," revision 3 had been issued July 29, 1985. This section instruction

letter is adequate to address the problem of maintenance

initiated couanon mode failures with two notable exceptions, t

( - 1. The potential exists for ccsunon mode failures to be

! caused by the use of the same calibrated tool on redundant pieces of equipment. While this is not specifically addressed in the NQAM except for reactor j protection and engineered safety features i

i 76

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

instrumentation, the potential for problems is significant, and NSRS recomunends that it be addressed in ISISL-A36. The Mechanical Maintenance Supervisor agreed.

2. The NQAM, part III, section 7.3, paragraph 3.3 states: "An important aspect of specific controls is redundancy -- redundancy of people, equipment, I inspections, review. . ." It also includes the following as an example of a specific control method 4

which may be employed: "The same individual should not be assigned to perform an identical activity on all similar units of multiple or redundant systems or i components." IstSL-A36 states that: " Supervisors l shall support the concept of maintenance redundancy, i i.e., redundancy of people, equipment, inspections, and review," and includes the example noted above from the WQAM. NSRS interprets this to mean that the i same individual should not perform the same

! maintenance activity on redundant equipment. The i practices of the electrical and instrumentation

! maintenance sections appear to be based on this interpretation.

i 3

Item 5 under Responsibilities in HMSL-A36 states:

" Assign Maintenance Request Planners the responsibility for identifying and indicating on the jl MR the possibility of conunon - mode problems prior to j placing CSSC MRs on the available status; i.e.,

4 similar maintenance on both RHR pumps should not be performed by the same person, or two qualified craftsmen should be assigned to the work." NSRS cannot agree that using two qualified craftsmen constitutes redundancy of people, and recommends that the phrase . . .'or two qualified craftsmen should be assigned to the work" be deleted.

l

NSRS also verified that foremen were being trained to l 19tSL-A36 and attended one of the training sessions. The 1 training was appropriate and pertinent and should increase I awareness of the potential for consnon mode failure. As a j possible improvement, MSRS suggests that a checklist of l

things to consider in addressing comanon mode failure be distributed to the trainees.

I This item remains open pending revision of MMSL-A36 to:

} (1) address the role of calibrated tools in potential j comanon mode failures, and (2) to meet the intent of

" redundancy of people" as stated in the WQAM.

RR. R-85-03-NPS-08, Surveillance of Maintenance program-Generic In the original review (reference A.54), NSRS determined I that surveillance of maintenance activities by onsite QA l

77 l

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, . . - , _..,.,....._m. , .~.,m.-, .,. . . , . . . _ . , . . , , - . , - - . . . . , _ . , - ~ , , m4... -, . .- . - , ,,-<,-

4 groups had not been adequately performed. NSRS recosumended , ,

that onsite QA groups perform indepth surveillances of the

maintenance program including review of items described in i other findings of the report. These items were concerned I with proper CSSC classification, post maintenance testing, ASME Section II testing, and consen mode failure.

The Nuclear Power response indicated that the Division of Quality Assurance (DQA) had issued a Management Review

' Guideline (MRG)-3.1, " Maintenance Performance" and planned  ;

to issue an MRC on postmaintenance test (PMT). The MRGs j

are intended to assist in achieving more indepth i

surveillances. The response also described the maintenance activities surveillances that had been performed.

For this follow-up review, MRG-3.1 and MRG-3.9, "Postmaintenance Testing," were reviewed and found to address all the items of concern except common mode I failure. The checklist for surveillance of PMT activities I had not been prepared and no PMT survalliances had been j scheduled.

h This item remains open pending: (1) addition of common mode failure to the surveillance program of maintenance activities (2) issuance of the PMT checklist, and i

(3) WSRS review of the implementation of surveillances on l the maintenance program and PMT.

! SS. New Recomunendation E-86-01-SQN-01, Increased Effectiveness l

of ALARA Program i

i For this follow-up review of I-84-12-SQN-13, it was I determined that sufficient corrective action has been taken I to satisfy the NSRS open item. However, discussions with j the HP staff ALARA engineer, and review of documentation 1ed the reviewer to conclude that additional specific l

l actions should be taken to strengthen the ALARA program at l SQN.

l Discussions with SQN personnel combined with infomation contained in each of the following documents was used to generate suggestions that can be used to determine What i

actions SQN will take: INPO Operational Experience Note

! REN/0EN-08A, "A Good Practice for the ALARA Program,"

Quality Audit Branch Audit Report No. QSS-A-85-0016 RC-14,

' " Radiation Work Pet 1mit (RWP) Program," and ROI-10

" Minimizing occupational Radiation Exposure." The following is a brief description of the pertinent-background inforination in each of the reports and the ,

I corresponding NSRS suggestion of possible actions SQN could take to improve the ALARA program.

. 78 4

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I i

1 j 1. Quality Audit Branch (OAB) *r ort No. OSS-A-85-0016 l

h QAB ALARA program audit was conducted at BFN and l SQN to verify that these plants have established and i implemented an effective ALARA program within the j scope of the quality assurance program to maintain low exposures. h deviations identified in the QAB

, report will not be addressed here; however, the ,

j observations and NSES suggestions are provided.

a. Staffina. The QAB report stated that BNF has i five senior health physics technicians assigned j

to specific ALARA activities working under the direct supervision of the ALARA engineer. During j the follow-up review, the SQN ALARA engineer said

{ that there is one full time HP technician j assigned to him; however, at the time of.this j follow-up review, he was supporting the ALARA

engineer on a half-time basis. The HP technican i attended the morning maintenance meetings to i provide ALARA input on HP, RWP shielding, and some ALARA preplanning. Based upon the current half-time effort, no coverage is being provided on backshifts, planning, modifications, or other j areas. There were also 104 ALARA preplans, and postplans prepared in 1985 and 12 in 1986 (as of I

February 21, 1986) which the ALARA engineer is required to approve. It is highly questionable that a single ALARA engineer and a half-time, or

{ even a full-time, technician can effectively accomplish the required tasks. The BFN staffing

{ 1evel of five full time HPs may not be the

! appropriate nsanber for SQN; however, more than i one is considered to be necessary. NSRS suggests

{ that SQN determine the appropriate HP staffing i level required to effectively perform the ALARA i duties during notinal and off-normal hours. 'This  !

! determination needs to consider all plant functions that require ALARA considerations; such  !

i as, maintenance, operations, modifications, outage planning, test, design, and site services. A job-task analysis could be used to

determine an effective staffing level.

-t i '

(

b. ALARA Review C - Ittee The QAB audit recomunended i i that SQN consider the merits and possible j benefits of an SQN ALARA review comunittee as is ,

currently operating at BFN. For this follow-up l review, it was determined that the INPO

} Operational Experience Note also recommends an

j. ALARA comunittee that has: (1) responsibility for

! overall coordination of the ALARA program, (2) be i composed of members .from the major functional

, 79 i

j I

i

,,m-. .-- . , - . . ,. _-, . - ~ - - , . . - _ - , , - . _ - , . _ _ . . . . _ _ , _ . , . . _ , , . . - , . . _ _ - . , . . . - - , + , - - , , . , . - - - . . , , , -

-w-,

, departments (3) meet on a regular basis to , ,

I review the status of the ALARA program, (4) and 4

review exposure reduction plans for specific jobs with estimates of 25 man-ren or greater.

The BF3 ALAaA review coRERittee, in ils third year i of operation, has the responsibility to review and direct the implementation of approved ALARA L suggestions. In addition the comunittee: reviews i planning schedules; discusses specific and timely ALARA problems; such as, reports of unnecessary

loitering in dose areas; reviews personnel

! contamination reports; reviews corrective action l

on delinquent postjob; ALARA reports; reviews j status of ALARA projects.

i j There was no evidence at SQN to suggest that any j of the above areas are effectively being j performed; e.g., the SQN 1985 ALARA goal was established late (March 1985) at 750 man-rem and

the actual was approximately 1100 man-rem. The

! ALARA engineer (nor others) was not aware of any critique that was perforined to attempt to improve 4

future perfotiaance and reduce doses. An ALARA connaittee would do so. Considering the apparent success at BFW and the INPO Good Practice '

recomunendation, WSRS suggests that the SQN plant consider establishing an ALARA review comunittee

composed of members from the major functional areas with the responsibility for overall l coordination of the ALARA program. Specific functions would include

(1) Review exposure reduction for specific jobs with exposure estimates greater than 25 man-rem.

(2) Direct the implementation of approved ALARA suggestions.

(3) Review planning schedules.

(4) Review specific and timely ALARA problems, such as, l reports of unnecessary loitering in dose areas.

l (5) Review personnel contamination reports, i

( (6) Review corrective action on delinquent postjob ALARA reports.

' i (7) Review status of ALARA projects. '

(8) Other.

! 80 l

l

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

The ALARA Committee composition and responsibilities could be incorporated into a plant instruction, e.g., an SQN Standard Practice or RCI.

c. ALARA Svanestion Program. The QAB audit determined that participation in the ALARA suggestion program has been poor at SQN. In the first year (1984) of the SQN ALARA Puggestion program, approximately 35 legitimate suggestions had been submitted. Less than ten were submitted by November of 1985. For this follow-up review, the ALARA engineer confirmed the lack of employee participation to the NSRS reviewer and also stated that no suggestions were received in 1986 as of February 21, 1986. The QAB audit recommended (and the NSRS concurs) that an employee award system (similar to BFN) be considered for SQN to stimulate additional employee involvement. Several award mechanisms identified were: day off with pay; savings bond; reserved parking spot; picture in plant newsletter; and ALARA T-Shirts, hats, or pens.

WSRS suggests SQN take action to increase employee participation in the ALARA suggestion program. Adoption of an awards program could be a way to increase participation,

d. ALARA Coordinators This item was not specifically addressed in the QAB audit report; however, the INPO Good practice reconumends the use of department ALARA coordinators. This is a staff position within the Radiological Protection Department with functional authority for implementation of the ALARA program and maintenance of the necessary records and data bases. For this follow-up review, discussions were held with the HP Section Supervisor Who stated that this type of function is being considered at SQN by assigning of an M-3 HP to assist the maintenance planners in ALARA preplanning and RWPs. This is a positive step and should be pursued. However, assignment of ALARA quellfied individuals to other groups (modifications, operations, test, site, services and design) should be made.

NSRS suggests that consideration be given to assigning that ALARA coordinators to all SQN site functional groups (modifications, maintenance, operations, test, design, site services, etc.) to provide these groups with the ALARA expertise to effectively implement the ALARA program.

Their function would be to assist in preplan preparation, postplan critique, dose reduction suggestions, maintain the necessary records and data bases, and incorporate industry experience into the SQN operations. This is considered to be an extension of the M-3 assignment mentioned previously.

e. Trainina. For this follow-up review, the SQN ALARA engineer stated that the work supervisors that propero ALARA pre and postplans, in many cases, cannot do an effective job. The HP section supervisor stated that an extensive ALARA an training program had been prepared but 81

never implemented. NSRS suggests that SQN consider

• preparing ALARA training program and that it be given to all individuals responsible for the ALARA effort, such as, the ALARA committee members, ALARA coordinators, and the individuals responsible for the preparation of the ALARA pre and postplans. This training program could include the fundamental principles of all radiation shielding and attenuation, and provide descriptive methods available to reduce dose levels in addition to time, distance, and shielding, e.g., changing test frequencies or times of test, preventive maintenance or design changes (such as moving high failure rate components or high frequency maintenance items out of radiation areas or provide permanent shielding), and cleaning or draining / refilling systems, etc.

V. LIST OF PERSONNEL CONTACTED A. Sequoyah Nuclear Plant Kathryn W. Allen, Reactor Engineer Larry D. Alexander, Mechanical Supervisor Field Services Group Ronald D. Bates, Mechanical Engineer Robert C. Birchell, Plant Compliance, O&PS Gary S. Boles, Mechanical Group Supervisor John C. Brady, Mechanical Engineer Mark E. Brock, Electrical Maintenance, POB i

Donna M. Bruno, Personnel Clerk I

Larry S. Bryant, Mechanical Maintenance Supervisor Marcia A. Cooper, Mechanical Testing, O&PS David L. Cowart, Quality Surveillance Supervisor Edward A. Craigge, Safety Supervisor, Industrial Safety Engineer i

Doug Craven, QA Supervisor Donald E. Crawley, Health Physics Supervisor

Don L. Deakins, Jr., Operations, POB John F. Denver, Mechanical Engineer Hugh D. Elkins, Jr., Instrument Maintenance Supervisor 82 m-- m --7, - , - - q p.,--, - - ,-,-u ~ - , , - - - - , - -

. a Steven V. Emert, Electrical Engineer Richard W. Farner, Instrument Engineer Ronald W. Fortenberry, Reactor Engineering Supervisor Timothy M. Galbreth, Employee Concern Program Site Representative Gary W. Gault, Reactor Engineering Supervisor John L. Hamilion, QA/QC Engineering Supervisor Philip R. Hitchcock, Mechanical Engineer Stephen P. Holdefer, HP Unit Supervisor, Support John F. Klein, Mechanical Engineer Tom D. Knight, Assistant to Site Director Tom Kontovich, Electrical Engineer Bennett C. Lake, Operations John A. Leamon HP ALARA Engineer Frank H. Lewis, QA Engineer Timothy E. Massey, Mechanical Engineer Mildred M. McGuire, Configuration Control Manager Manoj P. Mehta, Modifications Scheduling Supervisor Lawrence M. Nobles, Operations & Engineering Superintendent Robert W. Olson, Modifications Branch Manager Roger D. Poole Instrument Engineer David C. Queen, Mechanical Engineer Heyward R. Rogers, Compliance Engineer Roswell F. Schnur, Instrument Engineer Michael R. Sediseik, Electrical Modifications Supervisor Mark A. Skarzinski, Electrical Maintenance Supervisor Joseph S. Steigelman, Unit Supervisor, Operation John M. Stitt, QC Shift Supervisor Victor M. Taylor, Safety Specialist 83

Gary E. Tiner, Instrument Engineer ,

Philip R. Wallace, Plant Manager Patricia Wilson, Administrative Services B. Watts Bar Nuclear Plan _t Gerald Brantley, Employee Task Force Jerry Collins, Mechanical Maintenance Supervisor Ed Dudley, Maintenance Training Supervisor Edward Elam, Mechanical Engineer Craig F. Faulkner, Reactor Engineer Gary J. Johnson, Reactor Engineer Marvin K. Jones, Engineering Group Supervisor Samuel Lingenfelter, Acting Mechanical Engineering Supervisor Robert C. Manley, Planning Supervisor l

Charley Margraves, Preoperational Test Engineer C. Chattanooaa Douglas A. Bateson, Plant Training Officer

! Howard B. Burdette, Mechanical Maintenance Instructor, POTC Nuclear Service Frank Chicketto, Power Operations Training Center I John Fox, Supervisor of Welding and Metallurgical Section, Nuclear Service Robert M. Harris, Supervisor, Maintenance Training Unit POTC i

Ellen Hensley, Health Physics Technician i Charles E. Kent, Jr., Nuclear Services l

David Lambert Licensing Felix A. Szczepanski, Chief Nuclear Safety Staff D. Browns Ferry Nuclear P18nt Bill Nichols, Maintenance Training Supervisor 84

E. Bellefonte Nuclear Plant John Bynum, Mechanical Maintenance General Foreman Jay Krell, Maintenance Superintendent F. Muscle Shoals John L. Lobdell, Nuclear Services Gilbert F. Stone, Director of Occupational Health and Safety VI. REFERENCES A. IVA Memoranda

1. Memorandum from H. N. Culver to W. F. Willis dated March 13, 1980, "TVA Shift Technical Advisor Program-NSRS Inspection and Evaluation" (NSR 800313 050)

2. Memorandum from H. N. Culver to J. R. Calhoun dated June 27, 1980, "Sequoyah Nuclear Plant Unit 1 - NSRS Review Report No.

R-80-05-SQW" (GNS 800627 002)

3. Memorandum from H. N. Culver to J. R. Calhoun dated August 25, 1980, "Sequoyah Nuclear Plant Unit 1 - WSRS Review Report No.

R-80-11-SQN" (GNS 800826 002)

4. Memorandum from H. N. Culver to H. J. Green dated May 5, 1981, "Sequoyah Nuclear Plant - Nuclear Safety Review Staff Report No.

R-81-07-SQN" (GNS 810505 052)

5. Memorandum from H. J. Green to H. N. Culver dated June 27, 1981 "Sequoyah Nuclear Plant - Nuclear Safety Review Staff Report No.

R-81-07-SQN" (GNS 810623 101)

6. Memorandum from H. W. Culver to M. N. Sprouse dated June 28, 1982, "All Nuclear Plants - Realignment of Open Test Lines -

Nuclear Safety Review Staff Report No. R-82-04-NPS" (GNS 820629 050)

7. Memorandum from H. N. Culver to W. F. Willis dated September 9, 1982, "Sequoyah Nuclear Plant (SQN) - Field Monitoring Team's Performance During July 8-9, 1982 SQW Radiological Emergency Exercise - Nuclear Safety Review Staff (NSRS) Report No.

I-82-20-SQN" (GNS 820909 052)

8. Memorandum from H. G. Parris to W. F. Willis dated September 27, 1982, "Sequoyah Nuclear Plant (SQN) - Field Monitoring Team's Performance During July 8-9, 1982 SQW Radiological Emergency Exercise - Nuclear Safety Review Staff (NSRS) Report No.

I-82-20-SQN" (GNS 820930 103) 85

9. Memorandum from E. A. Belvin to H. N. Culver dated October 22, ,

1982, "Sequoyah Nuclear Plant (SQW) - Field Monitoring Team's Performance During July 8-9, 1982 SQN Radiological Emergency Exercise - Nuclear Safety Review Staff (E325) Report No.

I-82-20-SQN" (GNS 821025 100)

10. Memorandum from R. B. Maxwell to J. W. Hufham dated October 28, 1982, " Evaluation of Potassium Iodide (KI) Action Levels for Field Monitoring Teams"

11. Memorandum from H. N. Culver to H. G. Parris dated December 1, 1982, "Sequoyah Nuclear Plant Investigation of 10 REM Extremity Exposure - Nuclear Safety Review Staff (NSRS) Report No.

I-82-21-SQM" (GNS 821203 050)

12. Memorandum from E. A. Belvin to H. N. Culver dated January 4, 1983, "Sequoyah Nuclear Plant Investigation of 10 REM Extremity Exposure - Nuclear Safety Review Staff (NSRS) Report No.

I-82-21-SQN" (GNS 830106 100)

13. Memorandum from C. C. Mason to T. G. Campbell dated January 7, 1983 " Nuclear Safety Review Staff (NSRS) - Report No.

I-82-21-SQN" (L53 830107 948)

14. Memorandum from H. N. Culver to G. H. Kimmons dated January 31, 1983, "Sequoyah and Watts Bar Nuclear Plants - Realignment of l Test Lines - Nuclear Safety Review Staff Report No. R-82-04-NPS Follow-Up" (GNS 830202 051)

15. Memorandum from' H. G. Parris to H. N. Culver dated February 11, 1983, "Sequoyah Nuclear Plant Investigation of 10 REM Extremity Exposure - Nuclear Safety Review Staff (NSRS) Report No.

I-82-21-SQN" (GNS 830214 100)

16. Memorandum from T. G. Campbell to C. C. Mason dated February 14, 1983, " Nuclear Safety Review Staff (NSRS) - Report No.

I-82-21-SQN" (L47 830203 802)

17. Memorandum from H. N. Culver to H. G. Parris dated March 9, 1983, "Sequoyah Nuclear Plant Investigation of 10 REM Extremity Exposure - Nuclear Safety Review Staff (NSRS) Report No.

I-82-21-SQN" (GNS 830309 050)

18. Memorandum from H. N. Culver to H. G. Parris dated August 1, 1983, "Sequoyah Nuclear Plant Investigation of 10 REM Extremity Exposure, NSRS Report No. I-82-21-SQN" (CNS 830309 050)

19. Memorandum from E. A. Belvin to H. J. Green dated March 24, 1983 "Sequoyah Nuclear Plant Investigation of 10-rom Extremity Exposure - Nuclear Safety Review Staff (NSRS) Report No.

I-82-21-SQN" (GNS 830328 102) 86

i.

~

20. Memorandum from E. A. Belvin to H.'N. Culver dated March 29, 1- 1983, "Sequoyah Euclear Plant Investigation of 10-Rem Rxtremity Exposure - Nuclear Safety Review Staff (NSRS) Report No. '

, I-82-21-SQN" (GNS 830330 101) d j 21. Mescrandum from H. J. Green to E. A. Belvin dated March 31 j 1983, "Sequoyah Nuclear Plant Investigation of 10-Rem Extremity

Exposure - Nuclear Safety Review Staff (MSRS) Report No.

j I-82-21-SQN" (GNS 830404 100) a j 22. Memorandum from E. A. Belvin to H. N. Culver dated June 23,

1983, "Sequoyah Nuclear Plant Investigation of 10-Rom Extremity l l Exposure - Nuclear Safety Review Staff (NSES) Report No.

I-82-21-SQN" (GNS 830627 100) i i 23. Memorandum from H. G. Parris to H. N. Culver dated August 1, 1983, "Sequoyah Buclear Plant Investigation of 10-Rom Extremity Exposure - Buclear Safety Review Staff (NSRS) Report No.

j I-82-21-SQN" (GNS 830802 101)

24. Memorandum from P. R. Wallace to All Section Supervisors dated l July 3, 1984, "SNP-Unit 1 Cycle 2 Outage Critique Meeting" i (L53 840702 943) i i j 25. Memorandum, from H. E. Culver to J. P. Darling dated August 1 j 1984, "Sequoyah Nuclear Plant (SQN) - Nuclear Safety Review i Staff (NSRS Investigation of Unit 1 Incore Instrumentation l Thimble Tube Ejection Accident on Apell 19, 1984 - NSRS Report No. I-84-12-SQN" (GNS 840801 050)

26. Memorandum from C. C. Mason to L. M. Mills dated August 3, 1984,

, "SQNP - Proposed Technical Specification Change No. 111" j (L53 840803 914) a f 27. Memorandum from F. A. Szczepanski to H. G. Parris dated i September 7, 1984, " Meeting No. 68 of the Sequoyah Nuclear j Safety Review Board (S-NSRS)" (L42 840907 801)

I

28. Memorandum from J. P. Darling to H. N. Culver dated September 18, 1984, "Sequoyah Nuclear Plant-Nuclear Safety  ;

q Review Staff (NSRS) Investigation of Unit 1 Incore j Instrumentation Thimble Tube Ejection Accident on April 19, 1984-NSRS Report No. I-84-12-SQN" (553 840918 905) a i 29. Memorandum from H. N. Culver to J. P. Darling dated September 24,1984, "Sequoyah Nuclear Plant (SQN)-Nuclear Safety Review Staff (WSRS) Investigation of Unit 1 Incore i Instrumentation Thimble Tube Ejection Accident on April 19, *

] 1984-NCRS Report No. I-82-12-SQN" (GNS 840924 052)

?

1 30. Memorandum from C. E. Kent to D. E. Crawley and A. W. Sorrell j dated October 3, 1984 "Special Evaluation Report-Evaluation of 4 i Extremity and Multibedging Dosimetry" (L49 841004 800)

) '

87 1

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

l

) .

i 31. Memorandum from P. R. M2Llace io There listed dated October 3

• j 1984, "NUC PR Action T.tatrA item iting from NSRS Report on Thimble

! Tube Ejection Event (S?) 34100') 975)

32. Memorandum from L. C. Ellis to P. r'. Wallace dated October 24, l 1 1984, "Use of Workplace Hazard Assess 4ent In Job Safety  ;

Analysis" (LOS 841024 800) j 33. Memorandum from R. E. A11sup to Those listed dated October 25,

1984, "SQN - WSRS Investigation of Unit 1 Inctrementation

! Thimble Tube Ejection accident on April 19, 1984 - NSRS Report I No I-84-12-SQN" (353 841029 870)

34. Memorandum from P. R. Wallace to E. R. Ennis dated October 30, 1984 (553 841029 869)

35. Memorandum from R. N. Butler to R. R. Wallace dated November 5, 4 1984, " Review of Maintenance Request (MR) Process"

36. Memorandum from J. P. Darling to H. N. Culver dated December 5, 4

1984, "SQW - WSRS Investigation of Unit 1 Incore Instrumen-tation Thimble Tube Ejection Accident on April 19, 1984 - NSRS Report No. I-84-12-SQN" (800 841129 801)

(

i 37. Memorandum from M. N. Culver to W. F. Willis, dated December 10 j 1984, " Response to Board Comment - Sequoyah and Watts Bar i Nuclear Plants - Manufacturer-Identified Misapplica-tion of Swagelok Tube Fittings at Westinghouse Reactor Seal l Tables" (GNS 841210 050)

38. Memorandum from C. F. Stone to J. P. Darling dated December 11, 1984, "SQN - NSES Investigation of Unit 1 Incore Instrumen-tation Thimble Tube Ejection Accident on April 19, 1984 - NSRS Report No. I-84-12-SQN (LOO 841212 069) f 39. Memorandum from C. R. Brimer to R. E. Alsup dated December 27, l

1984, "SQN - NSRS Investigation of Unit 1 Instrumentation

Thimble Tube Ejection accident on April 19, 1984-NSRS Report No.

1-84-12-SQN" (801 841227 952) l 40. Memorandum from D. E. Crawley to L. M. Nobles dated January,

! 1985 SQN-Health Physics Section-Monthly Report" i

41. Memorandum from K. W. Whitt to J. P. Darling dated March 12,

1985, " Nuclear Safety Review Staff (NSRS) Report R-84-17-NPS-j Review of Procurement Practices and Procedures for Operating i Nuclear Power Plants" (Q01 850312 050) l

, 42. Memorandum from K. W. Whitt to J. P. Derling dated March 25,

{ 1985. "Sequoyah and Watts Bar Nuclear Plants-8pecial Review of l

Manufacturer-Identified Potential Misepplication of Swagelok l Tube Fittings at Westinghouse Reactor Seal Tables - NSRS Report j No. k-85-02-SQN/WBN" (QO1 850325 051) l l 88 i

I

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

j

, e s - 43. Memorandum from K. W. Whitt to W. F. Willis dated March 25, 1985, "Sequoyah and Watts Bar Nuclear Plants - Special Review of i

Manufacturer-Identified Potential Misapplication of Swagelok Tube Fittings at Westinghouse Reactor Seal Tables - NSRS Report No. R-85-02-SQN/WBN" (QO1 850325 050)

44. Memorandum from H. L. Abercrombie to J. W. Hufham dated

April 17,1985, "Sequoyah and Watts Bar Nuclear Plants -- Special ,

Review of Manufacturer Identified Potential Misapplication of Swagelok Tube Fittings at Westinghouse Reactor Seal Tables -

NSRS Report No. R-85-02-SQN/WBN" 1

(S53 850415 838)

I 45. Memorandum from TVA Board of Directors to All TVA Employees dated April 30, 1985, "TVA Policy on Reporting Nuclear Safety Matters" i

46. Memorandum from J. P. Darling to K. W. Whitt dated May 6,1985, "Sequoyah (SQW) and Watts Bar (WBN) Nuclear Plants - Special Review of Manufacturer - Identified Potential Misapplication of Swagelok Tube Fittings at Westinghouse Reactor Seal Tables-NSRS Report No. R-85-02-SQW/WBN" (L44 850426 807) 4

47. Memorandum from J. P. Darling to Those listed dated May 14, i 1985, " Procedure for Serious Incident Investigations" (L42 850429 801) 2 A8. Memorandum from G. B. Kirk to H. L. Abercrombie dated June 11, 1985, " Independent Safety Engineering Group (ISEG) Investiga-tion on Auxiliary Feedwater Recirculation Misalignment on April 16, 1985" (300 850611 800) r

49. Memorandum from G. E. Kirk to H. L. Abercrombie dated June 11, 1985, " Independent Safety Engineering Group (ISEG) Investiga-tion on Underlying Causes of Valve Misalignment on ESF Pump Room Coolers" (300 850611 801)

! 50. Memorandum from J. P. Darling to Those listed dated June 14,  :

} 1985, " Employee Concern Program," (LOO 850614 883) ,

51. Memorandum from W. T..Cottle to J. W. Hufman dated June 26,

! 1985, " Watts Bar Nuclear Plant (WBN) - Review of Maintenance j Program - Nuclear Safety Review Staff (NSRS) Report No.

l R-85-03-NPS" (T03 850626 830)

! 52. Memorandum for H. L. Abercrombie to J. W. Hufham dated June 28, l 1985, "Sequoyah Nuclear Plant (SQN) - Sequoyah, Browns Ferry.

and Watts Bar Nuclear Plants - Review of Maintenance Program -

Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NFS" (353 850626 913)

53. Memorandum from R. J. Mullin to J. W. Hufham dated July 2, 1985, "Sequoyah (SQN), Browns Ferry (BFN), and Watts Bar (WBN) Nuclear i

Plants - Review of Maintenance Program - Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NPS" (LO4 850701 818) 89

. +

54. Memorandum from K. W. Whitt to J. P. Darling dated July 5, ,

1985, "Sequoyah, Browns Ferry, and Watts Bar Nuclear Plants -

Review of Maintenance Program - Nuclece Safety Review Staff (NSRS) Report No. R-85-03-NPS" (QOl 850705 050)

55. Memorandum from J. P. Darling to K. W. Whitt dated July 11, 1985 "Sequoyah (SQN), Browns Ferry (BFN), and Watts Bar (WBN)

Nuclear Plants - Review of Maintenance Program - Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NPS" (L44 850702 808)

56. Memorandum from B. M. Patterson to R. M. Harris dated July 22, 1985, " Tube Fitting C?. ass"

57. Memorandum from J. P. Darling to K. W. Whitt dated July 25, 1985, "Sequoyah (SQN), Browns Ferry (BFN), and Watts Bar (WBN)

Nuclear Plants - Review of Maintenance Program - Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NPS" (L44 850718 802)

58. Memorandum from J. M. Anthony to SEs, ASEs, UOs and AUOs dated August 28, 1985, "SQN - Isolation of All Auxiliary Feedwater Recire Valves" (50 RIMS)

59. Memorandum from W. T. Cottle to Those listed dated September 20, 1985, " Tube Fitting Awareness and Initial Training" (L29 850909 817)

60. Memorandum from R. J. Johnson to W. T. Cottle dated Getober 9, 1985, " Tube Fitting Awareness and Initial Training" (L47 851009 800)

61. Memorandum from R. L. Lewis to T. F. Ziegler dated October 25, 1985. "BFN - Tube Fitting Awareness and Initial Training" (R36 851021 904)

62. Memorandum from K. W. Whitt to H. G. Parris dated November 22, 1985, "Sequoyah (SQN), Browns Ferry (BFN), and Watts Bar (WBN)

Nuclear Plants - Review of Maintenance Program - Response to Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NPS" (QO1 851122 051)

63. Memorandum from M. R. Harding to P. R. Wallace dated November 29, 1985 "CO2 Protected Areas at Sequoyah Nuclear Plant"

64. Memorandum from R. J. Mullin to K. W. Whitt dated December 19, 1985 "Sequoyah (SQN), Browns Ferry (BFN), and Watts Bar (WBN)

Nuclear Plants - Review of Maintenance Program - Nuclear Safety Review Staff (NSRS) Report No. R-85-03-NPS" (LO4 851217 859)

65. Memorandum from P. R. Wallace to Those listed dated December 19, 1985, " Technical Review of all SURVEILLANCE INSTRUCTIONS -

Schedule For Review of Surveillance Instructions (Instrument Maintenance Group)" ,

90

- - - - , . , . - - . . .--c _ _ - .,- , - - , - - . - . . - - - - . . , - , . - . - - .

e .

66. Memorandum from G. W. Killian to Those listed dated December 20, 1985, " Transmittal of Quality Audit Branch Audit Report No.

QSS-A-85-0016" (L17 851220 800)

67. Memorandum from J. A. Domer to H. R. Denton dated December 27, 1985, " Address Qupl fee concern Program"

68. Memorandum from G. W. Killian to H. L. Abercrombie dated December 30, 1985 " Evaluation of Corrective Action Taken -

Deviation Report closure" (L17 851230 809)

69. Memorandum from M. M. McGuire to H. L. Abercrombie dated January 8, 1986, " Program Plan for Conversion to Configuration Control Drawings" (L16 860108 809)

70. Memorandum from C. C. Mason to Those listed, dated January 23, 1986, "Use of Vendor Mancals" (L12 860122 801)

71. Memorandum from G. B. Kirk to H. L. Abercrombie dated January 29, 1986, " Independent Safety Engineering Group (ISEC)

Investigation on Loss of Spent Fuel Pit Level on December 18, 1985" (300 860129 800)

72. Memorandum from G. B. Kirk to H. L. Abercrombie dated January 30, 1986, " Independent Safety Engineering Group (ISEG)

Investigation of Unreliability of the Units 1 and 2 containment Hydrogen Analyzers (Monitors) of June 1984 and June 1985" (500 860130 800)

73. Memorandum from S. A. White to Those listed dated January 30, 1986, "New TVA Employee Concern Program" (L12 860130 802)

74. Memorandum from H. D. Elkins to PORC CHAIRMAN-SQN dated February 3,1986, "SQN-Instrument Maintenance Instructions (IM's)-Review in accordance with AI-4"

75. Memorandum from M. M. McGuire to H. L. Abercrombie dated February 5, 1986, "Sequoyah Nuclear Plant - Identification and Schedule of Startup Items of Program Plan for Conversion to Configuration Control Drawings" (S01 860205 929)

76. Memorandum from D. H. Tullis to B. M. Patterson dated February 11, 1986, "Sequoyah Nuclear Plant Maintenance Instruction (MI) Writer's Guide Summary Report"'

77. Memorandum from H. D. Elkins to PORC Chairman-SQN dated February 24, 1986, "SQN - Instrument Maintenance Instructions (IM's) - Review in Accordance with AI-4"

78. MemorandumfromB.M.PattersontoP.R.Wallaceda[ed February 26, 1986, " Review of Thimble Tube Maintenance l Instructions for Steps Which May Apply Excessive Forces to High j Pressure Fittings" l

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j B. Sequovah Nuclear Plant , ,

.i Ad=4=latrative Instructions

1. SNP Administrative Instruction AI-3, Revision 29, " Clearance Procedure," dated January 30, 1986.

2. SNP Administrative Instruction AI-5, Revision 35. " Shift and Relief Turnover" dated December 17, 1985.

3. SQN Administrative Instruction AI-8, " Access to Containment,"

Revision 16, approved January 13, 1986.

7

4. SQN Administrative Instruction AI-9, " Control of Temporary Alterations and Use of the Temporary Alteration Order,"

Revision 19, approved October 25, 1985.

SQN Administrative Instruction AI-19 (Part III). " Plant 5.

Modifications: Modification Requests," Revision 12, approved May 15, 1985.

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6. SQN Administrative Instruction AI-19 (Part IV), " Plant Modifications: After Licensing," Revision 13 approved January 23, 1986.

7. SQN Administrative Instruction AI-25 (Part I), " Drawing Control After Unit Licensing," Revision 12, approved January 30, 1986.

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! 8. SQN Administrative Instruction AI-25 (Part II), " Revision of As-Constructed Drawings," Revision 0, approved October 25, 1985.

9. SNP Adminstrative Instruction AI-27 Revision 7. " Shift Technical Advisor" dated September 24, 1985.

Annual Trend Analysis i

i 10. SQN " Annual Trend Analysis of Estimated Section Exposure" dated February 6, 1986.

, Chante 50.

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11. SQN Change No. 85-1569.

12. SQW Change No. 85-1599.

13. SQN Change No. 85-1699 to SQM-2 approved December 31, 1985.

I Electrical Maintenance

14. SQN Electrical Maintenance Section Instruction Letter EMSL-A36, I " Common-Mode Failures, Maintenance Initiated " Revision 1 approved April 7, 1976 l

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i Entineerina Section

15. SNP Engineering Section Instruction Letter ES SIL All, Revision 1 " Station Shift Technical Advisor Training" dated March 27, i 1985.

! General Operatina Instruction

16. SQW General Operating Instruction COI-01, " Plant Startup from-Cold Shutdown to Hot Standby," Revision 56, approved

! December 6, 1985. .

i Hazard Control Instruction

17. SQN HCI-M23. " Tube Fittup," dated November 8, 1985.

Hesith Physics t

O j 18. CQN Health Physics Section Instruction Letter HPSIL-1, i

" Radiation Surveys, Revision 12, approved May 14, 1985.

d j 19. SQN Health Physics Section Instruction Letter HPSIL-25, "ALARA j Program " Revision 3, approved March 22, 1985.

4 l 20. SQN Health Physics Section Instruction Letter HPSIL-27,

" Multiple TLD Badging," Revision 4, approved August 26, 1985.

21. SQN Health Physics Section Instruction Letter HPSIL-28, Attachment 1, " Quarterly Emergency Van Inventory " performed November 6, 1985, for van No. 10279, i  !

Imolementina Procedures Document i

22. SQNP Implementing Procedures' Document IP-20. " Environmental Monitoring During a Radiological Emergency," Revision 3 approved August 22, 1985.

Instrument Maintenance

23. SQN Instrument and Controls Training Continuing Pneumatic i

Valve Stroking SQN-INC-20, Revision 0. December 23, 1985.

I j 24. SQN Instrument Maintenance Instruction IMS-A4, " Independent Verification," Revision 0, i

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25. SQN Instrument Maintenance Section Instrudtion Letter 15S-132,

" Checklist Review of Plant Procedures," Revision 0, dated November 27, 1985.

Maintenance Instructions

26. SQN Maintenance Instruction Writer's Guide, DRAFT ONLY.

27. SQN Maintenance Instruction MI-1.9, " Bottom Mounted Instrumentation Thimble Tube Retraction and Reinsertion" Revision 7 dated September 9, 1985, and next revision DRAFT.

28. SQN Maintenance Instruction MI-1.10. "Incore Flux Thimble Cleaning and Lubrication," Revision 3, approved September 9, 1985, and next revision DRAFT.

29. SQN Maintenance Instruction MI-1.11. " Thimble Tube Installation," Revision O. DRAFT ONLY.

30. SQN Maintenance Instruction MI-6.20 " Configuration control During Maintenance Activities," Unit 0, Revision 6, approved November 26, 1984.

31. SQN Maintenance Instruction MI-6.24, " Inspection of High-Pressure Fire Protection Strainers," Revision 1, approved January 28, 1986.

32. SQN Maintenance Instruction MI 10.05.1, " Boric Acid Transfer Pumps, Revision 0, DRAFT ONLY.

Mechancial Maintenance

33. SQN Mechanical Maintenance Section Instruction Letter MMSL-A36,

" Common-Mode Failures, Maintenance Initiated," Revision 3, approved July 29,'1985.

34. SQN Mechanical Maintenance Section Instruction Letter MMSL-A65,

" Mechanical Craft Training Program," Revision 4, approved

! January 24, 1986, i

Monthly TACF Status Report

35. SQN Monthly TACF Status Report for January 1986,

36. SQN Monthly TACF Status Report for November 1985, i

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Ouality Assurance

37. SQN Quality Assurance Section Instruction Letter No. 5.1,

" Plant Instructions - QA Staff Review," Revision 5, dated October 18, 1985.

38. Quality Surveillance Section Annual Plan 1985.

39. QA Survey, Checklist No. 8a-85-P-001, Equipment Status, May 8-17, 1985, dated May 28, 1985.

40. QA Survey, Checklist No. 4a-85-A-005, Maintenance Activity Surveillance, April 26-17, 1985, May 7, 1985.

41. QA Survey, Checklist No. 20a-85-I'-002, TACFs, June 12-24, 1985, dated June 28, 1985.

42. QA Survey, Checklist No. Ic-85-S-004, Drawing Control - Unit 1, May 16-24, 1985, dated May 31, 1985.

43. QA Survey, checklist No. Ic-85-P-005, Revision 1, "As-Constructed" Drawing Verification, November 7-14, 1985, dated January 21, 1986.

Quality Enzineering

44. SQN Quality Engineering Section Instruction Letter No. 5.1,

" Plant Instructions - FQE Section Review," Revision 3, dated April 17, 1984. l Radiolozical Control Insttvetions

45. SQN Radiological Control Instruction RCI-1, " Radiological Hygiene Program," Revision 28, approved December 20, 1985.

46. SQN Radiological Control Instruction RCI-3, " Personnel Monitoring," Revision 22, approved January 13, 1986.

47. SQN Radiological Control Instruction RCI-10, Revision 8

" Minimizing Occupational Radiation Exposures," dated June 7, 1983.

48. SQN Radiological Control Instruction RCI-10, Revision 10

" Minimizing Occupational Radiation Exposures," dated December 11, 1985.

49. SQN Radiological Control Instruction RCI-14. " Radiation Work Per1 alt (RsJP) Program," Revision 4, approved July 20, 1985.

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Radiolonical Protection ,, ,

50. SQN Reason Plan ECP Instructor Notes.

l Special Maintenance Instruction

51. SQN Special Maintenance Instruction SMI-0-68-28, " Change-out of f the RCS Marrow Range RTDs," Revision 0, November 19, 1985.

52. SQN Special Maintenance Instruction SMI-0-94-3, " Seal Table High Pressure Seal Repair," Revision 1, approved January 17, 1986, and next revision DRAFT.

53. SQN Special Maintenance Instruction SMI-1-94-5, " Thimble Tube Installation " Revision 1, apprqved May 25, 1984.

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! Special Test Instruction

54. SQN Special Test Instruction SQ-STEAR-INST 82-12. " Turbine Benchmark Radioactive Tracer Test Unit 1," Revision 3, approved April 17, 1984.

Special Tool Evaluations

55. Special Tool Evaluation M-3-26-4, " Fire Pump Stand,"

October 3, 1985.

56. Special Tool Evaluation, M-3-62-7, " Centrifugal Charging Pump Element Handling Beam," October 3, 1985.

57. Special Tool Evaluation, M-3-68-12. "Conoseal Tool", August 1, 1985.

l 58. Special Tool Evaluation, M-3-68-13. "RPV Stud Elongation Rods j w/ ends Machined Flat," October 3, 1985.

59. Special Tool Evaluation, M-3-68-14 "RPV Stud Nut Cleaner,"

I October 3,'1985. t l 60. Special Tool Evaluation, M-3-68-15 "RPV Stud Cleaner,"

j October 3,1985.

61. Special Tool Evaluation, M-3-68-16. " Fast Stud Spinout Tool "

October 3, 1985.

l l 62. Special Tool Evaluation, M-3-68-17 " Steam Generator Sludge Lanceing Tools," October 3, 1985.

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63. Special Tool Evaluation, M-3-68-18 " Steam Generator Tube Plugging Tools " October 3, 1985.

64. Special Tool Evaluation, M-3-74-6 " Residual Heat Removal Pump Motor Lif ting Brackets," October 3,1985.

65. Special Tool Evaluation, M-3-68-1, " Reactor Coolant Pump #1 Seal Housing Centering Bracket," October 3, 1985.

66. Special Tool Evaluation, M-3-68-2, " Reactor Coolant Pump Seal Handling Rails Fabricated from Aluminum," October 3, 1985.

67. Special Tool Evaluation, M-3-68-3, " Reactor Coolant Pump Motor Shaft Centering Brackets," October 3, 1985.

68. Special Tool Evaluation M-3-68-8, " Reactor Coolant Pump Guide Studs for #1 Seal Housing Piece Fabricated from Aluminum,"

October 3, 1985.

69. Special Tool Evaluation M-3-68-10 " Reactor coolant Pump Coupling Puller," October 3, 1985.

70. Special Tool Evaluation, M-3-68-11. " Reactor Coolant Pump #1 Seal Housing Lifting Brackets," October 3,1985.

71. Special Tool Evaluation, M-3-82-5, " Turbocharger Lifting Beam,"

October 3, 1985.

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72. Special Tool Evaluation M-3-68-19 "RPV Head Stud Lift Rig,"

December 6, 1985.

73. Special Tool Evaluation, 86-1, " Lifting Eye for Containment Air Return Fan Motor," February 11, 1986.

Standard Practices

74. SQN Standard Practice SQA 129, " Objectives in Plant Operation-Sequoyah Nuclear Plant," Esvision 5, approved January 2, 1986.

75. SQN Standard Practice SQA 145," As Low As Reasonable Achievable (ALARA) Suggestion Program," Revision 0, approved January 19, 1986.

76. SQN Standard Practice SQA 166, " Program For Informing Employees How To Report Their Safety Concerns Revision 4, approved January 7, 1986.

77. SQN Standard Practice SQA 168, " Systems Engineering," Revision 0, approved January 10, 1986.

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78. SQN Standard Practice SQA 178, "TVA Office'of Nuclear Power ,

Employee Concern Program Line Organizations Procedure, Revision 0, dated February 6, 1986.

79. SQN Standard Practice SQM1, "Sequoyah Nuclear Plant Maintenance Program," Revision 0, approved January 18, 1983.

80. SQN Standard Practice SQM1, "Sequoyah Nuclear Plant Maintenance Program," Revision 4, approved January 8, 1986.

81. SQN Standard Practice SQM2, " Maintenance Management Program,"

Revision 16, approved December 27, 1985.

82. SQN Standard Practice SQM2, " Maintenance Management System,"

Revision 13, approved January 11, 1985.

83. SQN Standard Practice SQM63, "Special or Modified Tooling -

Primary Systems," approved May 9, 1985.

84. SQN Standard Practice SQS7 " Hazard Control Plan," Revision 2, approved January 7, 1986.

Startuo Instruction

85. SQN Startup Instruction SU-10.2 " Steam Generator Moisture Carryover Measurement," Revision 6, approved March 8,1982.

Surveillance Instnactions

86. SQN Surveillance Instruction SI-37, " Containment Spray Pump Test," Revision 16.

87. SQN Surveillance Instruction SI-146, " Reactor Coolant System Leak Test," Revision 12, approved May 1, 1984.

i 88. SQN Surveillance Instruction SI-250, " Reactor Coolant System l Hydrostatic Pressure Test," Revision 1, approved September 2, 1981.

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l Technical Instruction

, 89. SQN Technical Instruction TI-69, " Summary of Pre- and I Post-Maintenance Valve Tests for ABMW Section II and 10CFR50 Appendix J " Revision 9, approved October 28, 1985.

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Work /unintenance Roeuests

90. SQN MR A-539610 dated September 3,1985.

91. SQN MR A-563292 dated October 23, 1985.

92. SQN MR A-522137 dated November 29, 1985.

93. SQN WR 109473, MR A-532134 dated November 30, 1985.

94. SQN MR A-548350 dated December 5,1985.

95. SQN MR A-563137 dated December 18, 1985.

96. SQN MR A-563138 dated December 18, 1985.

97. SQN MR A-563139 dated December 18, 1985.

98. SQN WR 105405, MR A-561999 dated December 24, 1985.

99. SQN WR 105786 dated January 6, 1986.

100. SQN WR B-112780 dated January 7, 1986, 101. SQN WR 103937 dated January 9, 1986.

102. SQN WR 103941 dated January 9, 1986.

103. SQN WR 103942 dated January 9, 1986.

104. SQN WR 103945 dated January 9, 1986.

105. SQN WR 113377 dated January 14, 1986.

106. SQN WR 103945 dated January 14, 1986.

Workpisa 107. SQN Workplan 11878.

C. Office of Nuclear Power

1. Administrative Instruction Power and Engineering, "1I Expression of Employee Views," Revision 2, approved June 11, 1985.

2. NQAM, Part II, Section 6.4, " Control of Temporary Alterations " November 5,1984.

3. NQAM, Part III, Section 7.3, " Common-Mode Failures, Maintenance Initiated," dated January 15, 1981.

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4. Quality Notice, NQAM Part V Section 2.4, " Design Change

• Control System Using Design Change Supplements," effective date October 7, 1985 (L16 851007 806).

5. Nuclear Dispatch, "New TVA Nuclear Employee Concern Program I Place," dated February 14, 1986.

6. Office of Power Radiation Protection Plan Section A " Nuclear Power Plants " Revision 0, dated August 12, 1983.

7. Office of Power Radiological Protection Plan, Section A.

" Nuclear Power Plants, Revision 1, approved November 2, 1983.

8. Power and Engineering (Nuclear) Radiological Protection Plan Section A. " Nuclear Power Plants," Revision 2, dated December 6, 1985.

9. Procedure 0202.08 Revision 0, " Electrical and Mechanical Maintenance Craftsmen Training" dated January 6,1986.

10. REP-IPD, CECC-IP-9, " Emergency Radiological Monitoring Procedures," Revision 4 dated December 4,1985.

11. Management Review Guideline MRG-3.1, " Maintenance Performance," approved September 20, 1985.

l 12. Management Review Guideline MRG-3.9, "Postmaintenance Testing," approved December 16, 1985.

D. General Ersloyee Training

1. General Employee Training GET-2.1, " Health Physics Level 1 Training."

2. General Employee Training GET-2.2, " Health Physics Level II Training."

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3. General Employee Training GET-2.3, " Health Physics (Retraining)."

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4. General Employee Training GET-2.4, " Health Physics Level 0 Training" H. P. and Security BY-PASS Exams.

E. Regulatory

1. U. S. NRC Report of Inspection, 50-327/84-24, " Thimble Tube Ejection Event of April 19, 1984 Tennessee Valley Authority -

Sequoyah Nuclear Plant Unit 1" dated November 1984 Transmitted to TVA via Letter J. Nelson Grace to TVA dated March 7, 1985.

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l 2. U. S. NRC Report Nos. 50-327/85-45 and 50-328/85-45 dated February 18, 1986.

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3. U. S. NUREG/CR-1369, SAND 80-7054 Revision 1, " Procedures Evaluation Checklist for Maintenance, Test and Calibration Procedures used in Nuclear Power Plants."

F. Letters

1. Letter from L. M. Mills to Director of Nuclear Reactor Regulation dated October 2, 1984 " Proposed Technical Specifications - SQMP," (L44 841002 800).

2. Letter from P. R. Wallace to USNRC dated October 11, 1984,

" Tennessee Valley Authority - Sequoyah Nuclear Plant Unit 1 -

Docket No-50-327-Facility Operating License DPR-77-Reportable Occurrence Report SQRO-50-327/84030, Revision 1,"

(S53 841012 992).

3. Letter from Gilbert Commonwealth Engineers and Consultants to TVA dated October 30, 1985, transmitting a report " Assessment of the Design Control Program for the Sequoyah Nuclear Plant,"

(805 851031 004).

4. Letter from J. A. Domer to Director of Nuclear Reactor Regulation dated November 8, 1985." Proposed Technical Specification - SQNP." (L44 851112 803).

G. Watts Bar Nuclear Plant

1. Watts Bar Nuclear Plant Surveillance Instruction,

" Containment Spray Pump Test", SI-4.0.5.72-P, Revision 10, dated February 7, 1986.

2. W85 Maintenance Instruction MI-94.3, "Incore Flux Thimble Cleaning and Lubrication."

H. Industry

1. Heat Stress Management Program For Nuclear Plants, RPRI WP-4453, February 1986

2. INP0 85-026 dated June 1985. " Writing Guidelines Forin Maintenance Test and Calibration Procedures."

3. INPO Good Practice Operational Experience NOTE REN/0EN-08A "ALARA Planning for Station Work" dated September 1982.

4. Systematic Assessment of Licensee Performance, Inspection Report Numbers 50-327/85-22 and 50-328/85-22 Tennessee Valley Authority Sequoyah Units 1 and 2. March 1, 1984 through May 31, 1985, Attachment 2.

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5. Westinghouse Electric Corporation Trip Report dated March'26, 1985, "To observe The Flux Mapping Operation For Unit 1," * *

6. Workplan Hasard Assessment Manual, Division of occupational .

Health and Safety. September 1982.

I. Genersi

1. Commitment or Corrective Action Tracking Report No. 84148, dated October 2, 1984.

2. Comunitment or Corrective Action Tracking Report No. 84149, dated October 2, 1984.

3. Comunitment or Corrective Action Tracking Report No. 84153, dated October 2, 1984.

4. Cosunitment or' Corrective Action Tracking Report No. 84158, dated October 5, 1984.

5. Corporate Comunitment Tracking System NCO-85-0491-019, "17 new Maintenance instniction procedures to be issued June 30, 1986."

6. Corporate Comunitment Tracking System NCO-85-0491-017 "All maintenance instructions to be reviewed with checklists by ,

July 1987."

7. Final MR Review Class Training Record, February 6,1986.

8. Hazard Assessment Worksheets dated May 3, 1983. August 13, 1985, October 28, 1985, November 4, 1985, and November 14, 1985. ,

9. Initial MR Review Class Training Records, January 30, 1986.

10. Instruction Reviews (by craftemen) PMf 0920-030, 0921-030, 0921-030, 0233-031, and 0230-031. ,

11. Job Safety Anlysis dated November 10, 1984, " Entry into Pressuriser Enclosure with unit at 100% Power."

12. Lesson Program, "AI-3 Clearance Procedure Training / Retraining" Revision 2. dated May 24, 1985.

13. Maintenance Request monthly Report for January 1986.

14. Mechanical and Electrical SNP Clearance Procedure Examination (No date).

15. Official Notice Board Poster, "TVA Employee Concern Program."

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16. PORC Maintenance Instruction Subcossaittee Review Checklist.

17. Project U2 Force Target Schedule dated February 18, 1986.

18. Project U2 Force Working Schedule dated February 19, 1986.

.19. Reactor Rngineer Job Description Sc-4.

20. RWP for Reactor Building 02-1-85005, " Plugging Steam Generator Tubes Including All Related Support Activities. Manual Nethod" dated October 1, 1985.

21. Student Manual Nuclear Training Branch " Initial Tube Fitting Training", course NNT-28. Revision 1, dated October 8,1985.

22. TVA Employee Concern Program." Policy and Reporting Instruction."

23. Work Permit (RWP) Program," Revision 4, approved July 10, 1985.

24. SQN Training Attendance Records AI-14. RCI-10/ALARA Program, dated January 21, 1985; February 4, 1985; March 18, 1985; March 25, 1985; March 27, 1985; April 30, 1985; June 25, 1985.

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