ML18033B720
| ML18033B720 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/30/1991 |
| From: | Zeringue O TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9106050272 | |
| Download: ML18033B720 (36) | |
Text
ACCELERATED DISTRIBUTION DEMONSTPATION SYSTEM REGULATORY XNFORMATION DISTRIBUTION SYSTEM (RIDS)
ESSION NBR:9106050272 DOC.DATE: 91/05/30 NOTARIZED: NO
, DOCKET CIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH.NAME AUTHOR AFFILIATION ZERINGUE,O.J.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to NRC 910412 ltr re open items noted in Operational Readiness Assessment Team Insp Rept
'0-260/91-.201~.Corrective actions:Site Std Practice 2.2 developed re writing instructions
& writers guides.
DISTRIBUTXON CODE:
IE01D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response D
NOTES:1 Copy each to: S.Black,B.WILSON 05000260 RECIPIENT ID CODE/NAME HEBDON,F INTERNAL: ACRS AEOD/DEIIB DEDRO NRR SHANKMAN,S NRR/DOEA/OEAB EXTERNAL EG&G/BRYCEiJ
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NSIC NOTES:
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AEOD AEOD/TPAB NRR MORISSEAU,D NRR/DLPQ/LPEB10 NRR/DREP/PEPB9D NRR/DST/DIR SE2 NUDOCS-ABSTRACT OGC/HDS3 RGN2 FILE 01 NRC PDR COPIES LTTR ENCL 1
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NOTE TO ALL"RIDS" RECIPIENTS:
A D
D PLEASE HELP US TO REDUCE 4VASTE! CONTACT THE DOCUMENT CONTROL DESK, t
ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPXES REQUIRED:
LTTR 29 ENCL 29
Tennessee valley Authority. post ortice Bo>r 2000. Decatur,'Alabama 35609 O. J. 'Ike'eringue Vice President, Browns Ferry Operations NY30 55)
U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket Nos. 50-260 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 NRC INSPECTION REPORT 50-260/91-'201
RESPONSE
TO OPERATIONAL READINESS ASSESSMENT TEAN INSPECTION OPEN ITEMS This letter provides TVA's response to the letter from S. A. Varga to Dan A. Nauman dated April 12, 1991, which transmitted the subject report.
The report opened 13 items that identified weaknesses in several functional areas.
NRC has requested TVA to address the open items and provide a status of the corrective actions.
Enclosure 1 contains TVA's response to the open items.
Enclosure 2
contains a list of commitments contained in this letter.
As agreed with your staff, the due date for this response has been extended to May 31, 1991.
If you have any questions, please telephone Patrick P. Carier at (205) 729-3570.
Very truly yours, TENNESS E VALLEY AUTHORITY 0.;"
. Zeringue Enclosures cc:
See page 2
9l06050272 9l05::0 PDR ADOCK 05000260 A
0
U.S. Nuclear Regulatory Commission NAY 3 0 199) cc (Enclosures):
Ms. S.
C. Black, Deputy Director Project Directorate 11-4 U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike, Rockville, Maryland 20852 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637
- Athens, Alabama 35609-2000'r.
Thierry M. Ross, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Enclosure 1
Tennessee Valley Authority (1VA)
Browns Ferry Nuclear Plant (BFN)
Reply to Unit 2 Operational Readiness Assessment Team (ORAT) Inspection Inspection Report Number X6QL23MQ1 2
Establish administrative controls delineating responsibilities for communications between the Unit 2 and Unit 3 site organizations at management levels.
This item would be accomplished before restart.
A memorandum from H. F. McCluskey, Vice President, Browns Ferry Restart and
- 0. J. Zeringue, Vice President, Browns Ferry Operations to J.
R.
Bynum, Vice President, Nuclear Operations and D. E. Nunn, Vice President, Nuclear Projects dated April 18, 1991, delineates responsibilities and lines of communication between BPÃ Operations and BFN Restart Organizations.
BFN Operations is responsible for operations and programs at BFN which affect the operations and maintenance of the units.
BFN Restart under H. F. McCluskey (Unit 3) provides service to BFN Operations for implementing modifications to BFN Units 1 and 3
based on NRC commitments and established criteria in accordance with BFN procedures.
TVA provided the ORAT inspectors with a copy of this memorandum.
NRC agreed that with the memorandum this issue was closed on April 19, 1991, during the ORAT exit meeting.
Implement the requirements of amended Technical Specifications (TS) 6.5.1.2a to designate primary Plant Operations Review Committee (PORC) members, and to revise procedure Site Director's Standard Practice (SDSP) 27.4 accordin'gly.
This item would be accomplished before restart.
TVA submitted a TS change dated March 1, 1991, which identified PORC members based on the current site organization and titles.
The amendment to the TS has been approved by NRC and has been implemented.
NRC agreed that with the TS change this issue was closed on April 19, 1991, during the ORAT exit meeting.
t Additionally, SDSP 27.4, Plant Operations Review Committee, has been revised to better define PORC members and PORC alternates.
j l
4,
Page 2 of 10 2
-2 Complete an incident investigation and root cause analyses to evaluate management and personnel activities during a fuel handling activity while one source range detector may have been inoperable.
This item would be accomplished before restart.
TVA has completed an incident investigation that addresses the-points raised by the NRC.
Additionally, TVA performed a Human Performance Enhancement System evaluation and incorporated the results into the incident investigation.
TVA considers this issue closed.
Complete an incident investigation to determine the cause for inadequate management awareness and lack of adequate guidance for the protection of safety-related equipment against contaminants generated by paint removal.
This item would be accomplished before restart.
t 2
TVA has completed an incident investigation to address the concerns regarding the protection of safety-related equipment from contaminants generated by paint removals NRC agreed that this issue was closed on April 19,
- 1991, during the ORAT exit meeting.
Correct procedural deficiencies identified by ORAT, including procedure style guide terminology and definitions.
A.
Discrepancies were observed between the format and style of the procedures and the guidelines for preparing procedures, Plant Manager Instruction (PMI) 2.3, "Style Guide for Writing Instructions" (Revision 7).
Many of these discrepancies were due to a lack of clearly-defined criteria in the style guide and a lack of verification of the written procedures against the style guide (e.g.,
logic terminology, referencing and branching, emphasis techniques, and definitions of terminology).
B.
Operating Procedure O-OI-32, "Control Air System Operating Instruction" (Revision 12), was not correctly revised to reflect of an engineering design change (DCN W13656a).
Procedural Step 5.1.9.6 states "Check compressor FINAL STAGE DISCHARGE PRESS reads between 85 and 97.5 psig" instead of between 100 and 115 psig as required by the design change.
The team observed the four compressor units operating between 105 and 110 psig.
The licensee initiated corrective action to revise the procedural discrepancy.
4 e
Page 3" of 10 C.
Operating Procedure 2-0I-71, "Reactor Core Isolation Coolant Operating Instruction" (Revision 13), incorrectly stated that reactor core isolation coolant (RCIC) pump flow should be greater than 60 (rather than) 600 gpm.
D.
Operating Procedure O-OI-57D, "D.C. Electrical System Operating Instructions" (Revision 12), Step 5.1.4 required verification of battery room temperature between 57oF and 97oF.
There was no temperature indication observed in the battery room.
Additionally, Step 5.1.5 mentioned the use of a portable smoker and reach rod to detect air movement.
The equipment was not staged or referenced in the procedure preceding the step requiring the action.
E.
Operating Procedure 2-0I-74, "Residual Heat Removal System Operating Instruction" (Revision 18), Step 8.1.1.3, required that the residual heat removal (RHR) pumps be vented "until a solid stream of water is observed" at the pump vent.
The vent piping entered a closed drain in such a manner that water flow cannot be observed.
Additionally, Steps 8.8.11.5 and 8.8.11.6 directed the operator to proceed to apparently wrong steps of the procedure.
F.
There were no documented definitions and training on the terms used to describe system capability and status in the procedures (e.g.
"functional," "in service," "available to support," "available to supply," and "in operation support").
Operators could not discriminate among these terms.
Such lack of clarity had the potential to lead to variance in the performance and validation of associated procedures.
The team noted that while the above discrepancies were minor, they were indicative of a lack of attention to detail during the procedure verification process.
TVA has prepared a Site Standard Practice (SSP) 2.2 on writing procedures.
SSP 2.2 combines the style guide for writing instructions (PMI 2.3) and site writers guides into one document.
SSP 2.2 is in the process of being reviewed by the site organizations prior to implementation.
Additionally, guidance is provided in this SSP on logic terms, referencing and branching, emphasis techniques, and definitions of terms.
~tm B
Procedural provide an identified compressor Step 5.1.9.6 in 0-OI-32 has been corrected in Revision 13 to operational range which reflects the design pressures in DCN -13656a.
Procedural Step 5.1.9.6 now states "Check FINAL STAGE DISCHARGE PRESS reads between 90 and 110 psig."
4
Page 4 of 10 In 2-0I-71, Section 6.1, Step 6.1.2 states "Maintain RCIC flow greater than or equal to 60 gpm on RCIC SYSTEM FLOW controller, 2-FIC-71-36A."
This step is correctly worded to ensure that RCIC minimum flow will be maintained above the established minimum flow value of greater that 60 gpm while allowing Unit Operator flexibilityto set
~
controller at required flows for RCIC operation.
In the Precautions and Limitations section, Step 3.4 states "In the presence of a RCIC initiation signal, the RCIC PUMP MINIMUM FLOW VLV, 2-FCV-71-34, will open when system flow is less than 60 gpm and close when flow is greater than 120 gpm.
The valves ~U 'Quent auto open on low flow if an initiation signal is not present."
It should also be noted that in Section 4.0, DI Step 4.3, RCIC SYSTEM FLOW controller is in AUTO and SET at 600 gpm; therefore, no change is required.
There is no change required for the discrepancy identified concerning the temperature indications in 0-OI-57D.
The operator has calibrated pyrometers readily available to him with air probes or thermometers from the Tool Room.
Step 5.1.5 has been revised to change the requirements for a portable smoker and a reach rod and use instead
'smoke generation tubes/bulbs and a ladder.
A requisition number allowing easy acquisition of the smoke generation tubes has been included in the step.
The problem concerning Step 8.1.1.3 in 2-OI-74 has been resolved in Revision 19 by the addition of a precaution note in Step 3.39 which states that the venting process be established for approximately one minute for closed loop piping.
Additionally, a note has been placed above the vent valves identified in Step 8.1.1.3 which reads "For closed loop vents, vent for 1 minute."
Additionally, Steps 8.8.11.5 and 8.8.11.6 have now been revised to reference the correct procedural steps.
Procedures which support Unit 2, such as Operating Instructions (OIs) and Surveillance Instructions (SIs), clearly support the TS definition of operable.
The unique situation at BFN (with all three reactors defueled) reduced the number of plant systems necessary to comply with TS requirements.
The System Pre-Operability Checklist (SPOC) process was created to methodically return these systems to the point where they meet the requirements of the TS operability.
By procedure and training, the licensed operations personnel are the only ones who can declare the systems operable.
Because of the regulatory connotations associated with the term "operable," it could not be used on systems prior to the completion of the SPOC process.
Other terms such as "functionally," "in
Page 5 of 10 service," "available to support," "available to supply," and "in operation support" were created to uniquely distinguish performance of these systems when not required by TS.
When the systems on Unit 2 have been returned to service, only the terms "operable" and "inoperable" will be applicable for TS systems.
TVA considers this issue closed.
Since the discrepancies on operating procedures indicate a lack of attention'o detail during the procedure verification process, a memorandum has been issued from the Unit Operations Manager supervising the Operations procedures group to impress upon personnel their responsibilities when reviewing procedures.
The memorandum reminded the procedures group of the importance of taking time when verifying procedure revisions.
Even though the discrepancies identified by ORAT were minor, extra care should be taken not only to identify major technical problems, but typographical errors and misspelled words as well.
Include in the applicable power ascension test procedures functional organization description; shift staffing; review of procedural references prior to test performance; and independent qualified review of test result packages.
This item would be accomplished before restart.
Organization and shift staffing charts including the appropriate descriptive t
verbiage have been added to PMI 26.1.
These changes have been approved by the Joint Test Group (JTG).
TVA has performed a final review of the test procedures associated with the Power Ascension Test Program.
This review has confirmed that no procedural reference discrepancies exist.
PMI 17.1, Paragraph 4.12 specifies the requirement for an independent review of test results.
This individual performing this review shall be someone other that the test performer and 'will evaluate the technical acceptability of the test results and ensure that the test requirements have been satisfied.
NRC agreed that with these changes to the power ascension test procedures, this issue was closed on April 19, 1991, during the ORAT exit meeting.
2 Review all procedures required for power ascension and correct technical deficiencies identified by ORAT.
This item would be accomplished before restart.
A review of selected power ascension test procedures performed during the ORAT inspection revealed minor technical deficiencies.
A final review of test procedures associated with the Power Ascension Test Program has now been completed.
Technical deficiencies identified by ORAT have been corrected.
All power ascension test procedures have been approved by the JTG.
~
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Page 6 of 10 Subsequent to the ORAT inspection, TVA identified several areas in the Special Operating Instruction (SOI) that could lead to confusion of the procedure requirements.
A special TVA team has been formed to review the SOI prior to its implementation at each management assessment point during power ascension testing.
TVA has identified the systems and components in Unit 1 and 3 that are required to be functional to support the Units 1 and 3 systems required to be operable.
TVA should formally identify the specific Unit 1 and 3 components that need to be functional and define for these components the requirements to meet functionality.
This item would be accomplished before restart.
t 2
As discussed with the NRC inspectors during the inspection, TVA does not have a list of Unit 1 and 3 system components that must be functional to support the Units 1 and 3 systems required to be operable.
System components in Units 1 and 3 which are required to be operable to support Unit 2 operation are identified on the-Unit 1 and 3 separation drawings.
These drawings depict the interfaces between Units 1, 2, and 3.
Units~1 and 3 equipment required for Unit 2 operation are identified by color coding on these drawings.
These components will be operable to support Unit 2.
A formal, structural process is utilized for the evaluation of BFN work activities.
When surveillances or work activities are scheduled to be performed, the operational impact of those activities is evaluated by the Operations Work Control group.
Impact evaluations are completed by an onshift SOS to ensure someone knowledgeable of the current plant conditions reviews the work activities prior to their initiation.
During this process, should any system or component be identified as being inadequate in supporting the desired activity, re-evaluation of the task will be performed.
The methodology presently used by BFN works well.
TVA considers this issue closed.
2
-2 I
Conduct additional training for plant personnel on the locked valve program.
NRC inspector's concern was on additional training for operators.
This was confirmed in a telephone call on May 15, 1991, with Don Norkins of NRC.
Operations supported by Operations Training conducted live-time training with the shift crews.
This training instructed Operations personnel on those operating practices which require strengthening and at the same time stressed their importance.
Items addressed in the training included the locked-valve
- program, independent verification versus second party verification, and surveillance and procedural adherence.
This training has been completed.
TVA considers this issue closed.
0 I
1 4
Page 7 of 10 Walkdown of procedure 1-SI-4.5.B.ll revealed several weaknesses with the procedure.
A.
The "Caution" following Step 7.8.9 discussed appropriate compensatory measures for 10 Part CFR 50, Appendix R requirements without stating or referencing these requirements.
B.
The location information was inadequate.
The licensed operator was provided inadequate information for locating certain components.
Even after the operator telephoned the control room for assistance, some components still could not be located.
C.
With regard to Step 7.10 (Note 2), the operator and on-duty Shift Technical Advisor were unsure against which unit the resultant limiting condition for operation (LCO) would be tracked.
D.
The "Caution" following Step 7.10.12.3 did not contain the value for maximum current, and the caution was required before the pump start (Step 7.10.7).
E.
During questioning regarding independent verification of valves with control room indication, the operator was unaware of requirements in Procedure SDSP 3.15, "Independent Verification," Step 6.1.7, regarding use of different remote indicators, and of performing one check locally.
Conduct additional training for plant personnel on the implementation of the independent verification program.
t 2
Concerning the lack of specific compensatory measures for Appendix R when performing Step 7.8.9, the "Caution" has been revised to delete the requirement for Appendix R compensatory measures.
/
Jtm~B The description of component locations in the SI is consistent with the descriptions given in the OI checklist (e.g.,
OI valve checklist).
The licensed personnel's unfamiliarity with certain components (e.g.
manual valves) was due in large part to the SI being performed (validation performance) for the first time.
Step 7.10 (Note 2) of 1-SI-4.5.B.ll, Residual Heat Removal Unit Cross-tie for Unit 2 Operability, identifies a
LCO may have to be entered if the LCO time limit cannot be met.
Since Unit 2 is the only unit which requires this cross-tie to be operable at this time, the LCO would be tracked on Unit 2.
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Page 8 of 10 A specific value for maximum motor current is addressed within the procedure in the Precautions and Limitations section under Step 3.2.
The operator has to review these precautions and limitations prior to the performance of the pump testing steps, and this review is documented in Step 7.3.1.
t m E The response for Item E is discussed in the response to 260/91-201-09.
TVA considers this issue closed.
~
2 Field verify Rosemount transmitters for adequate cover torque, and evaluate the adequacy of the calibration procedure to assure the proper torquing of all transmitters.
Field verification of Rosemount transmitters for adequate cover torque has been completed.
TVA has also completed an additional evaluation of the calibration procedure and considers the process for the installation an'd torquing of Rosemount transmitters to be adequate.
Accordingly, TVA considers 2
2 Evaluate the weaknesses identified in the trending programs and implement appropriate corrective actions.
A.
The baseline 'for determining an adverse trend for several indicators was too high.
For example, a baseline of 50 percent for engineering calculations meant that adverse trends would only be identified when the reject rate for calculations exceeded 50 percent.
B.
In the monthly tracking of reject rates, for any month that a trend data audit was not con'ducted, the reject rate would be graphically indicated as zero.
For example, for four months between June and December
- 1990, the graph indicated a reject rate of zero.
The team determined that no audits were performed for three months during that
- period, thus providing erroneous reject rate data to management.
C.
Two Conditions Adverse to Quality items, drawing discrepancies, and work requests were identified by ORAT as not having trend data forwarded to site Quality Assurance (QA) as required by site QA.
n t
2 2 The control limits have been eliminated.
Instead of utilizing control limits as a threshold for management attention, each item trended is
I
Page 9 of 10 analyzed on its,own significance, merit, and impact.
Additionally, meetings are held with line organizations before the trend report is generated to discuss each problem area.
In these meetings the line organizations are made aware of each potential problem area so timely corrective actions can be initiated.
By utilizing this approach, management will be alerted to items that are significant and will produce the most impact on the plant.
During periods when trend data audits are not conducted, no trend data will be illustrated and therefore, will not be added into the overall trend analysis.
Site QA has contacted the line organizations and is now receiving the necessary trend data.
2
-2 Implement corrective actions to address weaknesses identified in the documentation of incident investigation reports.
A.
There was no indication in the incident investigation reports (IIRs) as to which one of the four identified root cause analysis (RCA) methodologies were used in the investigation.
B.
One investigation team member was required to be trained in RCA methods.
Since the team composition was not evident on the IIRs, the team was unable to determine who participated in the RCA and whether that member had been trained in RCA methods.
C.
Although Procedure SDSP 15.4 provided a suggested report format, those preparing IIRs did not regularly follow the procedure.
This lack of a formal report structure made it difficult to ascertain if all pertinent information had been supplied, including the proper identification of corrective action.
For instance, IIR 90-121 concerned a Unit 3 RHR pump motor heater cable with insulation damage.
A recommended action was to investigate the cable of all 4-kV motors on site for similar problems, however, this action was not identified in a corrective action section and thus was not appropriately followed through to completion.
D.
The team also identified two instances on IIRs90-014 and 90-146 where a corrective action identified in the IIR was altered during entry into the tracking system.
The program currently required that such changes be approved by the Plant Manager, such approval was not obtained.
0
Page 10 of 10 t
2
-2 Incident Investigations (IIs) can utilize a combination of RCA methods simultaneously and often use abridged versions as individual circumstances deem appropriate.
TVA emphasizes providing RCA training to personnel to ensure a basic understanding on root cause approach more than the mechanics of specific methods.
TVA will make appropriate changes to the IIR (i.e. Final Event Report) package format to facilitate identification of team members who have been trained in RCA methods.
A new corporate standard on IIs was recently approved and is in the process of being implemented as a site instruction.
This new procedure uses Institute of Nuclear Power Operation forms for both Category 1 and 2
event reports.
Use of common forms will promote uniform report presentation and format.
To address the ORAT comment on IIR 90-121, a specific tracking item was added to make the associated follow-up activities part of the IIR corrective actions.
~tg D
Regarding IIR 90-014, the original corrective action was to insulate penetrations in an outside pump house.
Fire Protection subsequently walked down the area in more detail and determined that no obvious benefit would result from sealing the building since a space heater was in the building.
The difference in the committed and actual corrective actions was later noted, and the Technical Support Manager approved the change.
The event report was a Category 4 (lowest significance report) and was originally approved by Technical Support.
A letter was put in the IIR file to clarify the paperwork sequence on this II.
The current II/RCA procedure uses only two report categories and requires Plant Manager approval of changes to committed actions.
Regarding IIR 90-146, the text of the corrective action statement stated in one instance that a particular procedure would be reviewed to determine if a procedure revision was necessary.
In the same paragraph, the stated corrective action was to revise the subject procedure.
To address the ORAT comment, the more specific action has been entered in the computer.
0702k
Enclosure 2
Tennessee Valley Authority (TVA)
Browns Ferry Nuclear Plant (BFN)
Reply to Unit 2 Operational Readiness Assessment Team (ORAT) Inspection Inspection Report Number Zfd923=@0, LIST OF COMMITMENTS 1.
Changes will be made to the Final Event Report package format to facilitate identification of team members who have been trained in root cause analysis methods by July 15, 1991.
0702k/17
0
h" SAq REGII.
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+**+~
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 April 12, 1991 Docket No.: 50-260 Hr. Dan A. Nauman Senior Vice President, Nuclear Power Tennessee Valley Authority 6N 38A Lookout Place 1101 Harket Street Chattanooga, Tennessee 37402-2801
Dear Hr. Nauman:
SUBJECT:
NRC INSPECTION REPORT:-50=260/91-201~BROWNS FERRY UNIT 2 OPERATIONAL READINESS ASSESSHENT TEAM INSPECTION This letter transmits the findings of the Operational Readiness Assessment Team (ORAT) inspection conducted during the periods February 11-15 and February 25 through Harch 1, 1991 involving activities authorized by NRC Operating License Number DPR-52 for the Browns Ferry Nuclear Plant Unit 2.
The inspection was conducted by members of the NRC Office of Nuclear Reactor Regulation, Office for Analysis and Evaluation of Operational
- Data, Regions I, II and IV, and NRC consultants.
0 The purpose of the team inspection was to perform an independent in-depth assessment of the readiness of management
- controls, programs, and personnel to support safe restart and operation of the facility.
The inspection team performed an integrated evaluation of various functional areas including facility management, operations, maintenance and surveillance, engineering and technical
- support, separation of operating and non-operating units, power ascension test program and quality verification.
Within these
- areas, the inspection consisted of interviews with personnel, observation of plant activi-
- ties, and review of procedures in the simulator and in the field.
The findings of the inspection team were discussed with you and other members of your staff at the conclusion of the inspection on Harch 1, 1991.
The inspection team identified four concerns which require resolution before startup.
A letter summarizing these concerns vras sent to you on Harch 15, 1991, in advance of the inspection report.
These concerns are also identified in the Executive Summary of the enclosed report.
Positive findings and weaknesses are characterized under each functional area.
Weaknesses that require additional actions on your part are identified as open items in the fuIIctional areas and are summarized in Section 10.0 of the report.
Please inform us in writing within 30 days as to the status of your corrective actions to address the open items.
>t(
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Apri1 12, 1991 The team did not identify any issues that would prevent them from recommending restart of Unit 2.
- However, due to the plant status during the inspection period, the team did not have adequate opportunity to observe activities in an operational environment.
The few activities that were evaluated, specifically refueling and paint removal, did not demonstrate to the team that all your activities would be conducted in a conservative manner.
Therefore, the team will return before restart to continue to assess your control over the conduct of activities that affect plant operations.
During this inspection, the team will also review your corrective actions in addressing the open items indicated in the attached report.
In accordance with 10 CFR 2.790 the Commission's regulations, a copy of this letter and the enclosures noted below will be placed in the NRC Public Document Room.
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
ORIGINAL SIGNED 'BY PNME A. VARGA
Enclosure:
Inspection Report 50-260 cc:
See page 3
Distribution: See page 4
Steven A. Varga, Director Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
- See previous concurrence RSI 8: DRIS RSI B: DR IS PSKol tay*:sf ftAtlil 1er 04/10/91 04/
/91 RSIB:DRIS FXTalbot 04/
/91 AEOD JVKauffman 04/
/91 SRI: RI V WRBennett 04/
/91 SRI: RI I RCButcher 04/
/91 AC: RSI B: DRIS EVImbro*
04/10/91 RI RJUrban 04/
/91 D:DRIS BKGrimes*
04/10/91 LHFB:DLPQ JAArildsen 04/
/91 04 91 LHFB:DLPQ GSGalletti 04/
/91 SC:RSIB:DRIS DPNorkin*
04/10/91
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Mr. Dan A. Nauman Browns Ferry Nuclear Power Station CC:
Mr. Marvin Runyon, Chairman Tennessee Valley Authority ET 12A 7A 400 West Summit Hill Drive
. Knoxvilie, Tennessee 37902 Mr. Edward G. Wallace
- Manager, Nuclear Licensing and Regulatory Affairs Tennessee Valley Authority 5N 157B Lookout Place Chattanooga, Tennessee 37402-2801 Mr. John B. Waters, Director Tennessee Valley Authority ET 12A 9A 400 West Summit Hill Drive
~ Knoxville, Tennessee 37902 Mr. Oliver D. Kingsley, Jr.
President, Generating Group Tennessee Valley Authority 6'8A Lookout Place 1101 Market Street Chattanooga, Tennessee 37902-2801 General Counsel Tennessee Valley Authority ET 11B 33H 400 West Summit Hill Drive Knoxville, Tennessee 37902 Mr. Dwight Nunn Vice President, Nuclear Projects Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801 Dr. Mark 0. Medford Vice President, Nuclear Assurance, Licensing and Fuels Tennessee Valley Authority 6N 38A Lookout Place Chattanooga, Tennessee 37402-2801 Mr. 0. J. Zeringue, Site Director Browns Ferry Nuclear Power Station Tennessee Valley Authority P. 0.
Box 2000
- Decatur, Alabama 35602 Mr. P. Carier, Site Licensing Manager Browns Ferry Nuclear Power Station Tennessee Valley Authority P.
O.
Box 2000
- Decatur, Alabama 35602 Mr. L.
W. Myers, Plant Manager Browns Ferry Nuclear Power Station Tennessee Valley Authority P. 0.
Box 2000
- Decatur, Alabama 35602
- Chairman, Limestone County Commission P. 0.
Box 188
- Athens, Alabama 35611 Claude Earl Fox, M.D.
State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W.
Atlanta, Georgia 30323 Mr. Charles Patterson Senior Resident Inspector Browns Ferry Nuclear Power Station V.S. Nuclear Regulatory Commission Route 12, Box 637
- Athens, Alabama 35611 Tennessee Valley Authority Roc kvi 1 1 e Office 11921 Rockvil le Pike Suite 402 Rockvi lie, Maryland 20852
1 4
t
Mr. Dan A. Nauman
~
~
Distribution:
50-260.
RSI8 R/F DRIS R/F
- SAVarga, NRR
- BKGrimes, NRR
- DPNorkin, NRR
- PSKoltay, NRR
- Director, DRP
- FJHebdon, NRR
- TMRoss, ERR
- MHKrebs, NRR
, Sr. Resident Inspector Regional Administrators Regional Division Directors Inspection Team LPDR PDR ACRS (3)
OGC (3) 1S Distribution
1 f
4 I