ML040760730

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Revised Minutes of Internal Meeting of the Davis-Besse Oversight Panel (Revised RAM Closure Attachment)
ML040760730
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 03/17/2004
From: Grobe J
NRC/RGN-III
To:
NRC/RGN-III
References
Download: ML040760730 (16)
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Text

March 17, 2004 MEMORANDUM TO: Davis-Besse Oversight Panel FROM: John A. Grobe, Chairman, Davis-Besse Oversight Panel /RA/

SUBJECT:

REVISED DECEMBER 23, 2003 MINUTES OF INTERNAL MEETING OF THE DAVIS-BESSE OVERSIGHT PANEL (Revised RAM Closure Attachment)

The implementation of the IMC 0350 process for the Davis-Besse Nuclear Power Station was announced on April 29, 2002. An internal panel meeting was held on December 23, 2003. Attached for your information are the minutes from the internal meeting of the Davis-Besse Oversight Panel, Inspection Plan Heat-up and Mode Change Inspection, Open Action Items List, and RAM Closure Forms.

Attachments: As stated cc w/att: D. Weaver, OEDO J. Caldwell, RIII G. Grant, RIII S. Reynolds, DRP R. Gardner, DRS B. Clayton, EICS G. Wright, DRP DB0350 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML040760730.wpd To receive a copy of this document, indicate in the box:"C" = Copy without enclosure "E"= Copy with enclosure"N"= No copy OFFICE RIII RIII RIII NAME RBaker CLipa JGrobe DATE 03/11/04 03/15/04 03/16/04 OFFICIAL RECORD COPY

MEETING MINUTES: Internal IMC 0350 Oversight Panel Meeting Davis-Besse Nuclear Power Station DATE: December 23, 2003 TIME: 1:30 p.m. Central ATTENDEES:

A. Mendiola B. Ruland J. Hopkins M. Phillips J. Stang J. Rutkowski S. Unikewicz R. Lickus J. Grobe D. Passehl M. Salter-Williams Agenda Items:

1. Discuss/Approve Todays Agenda The Panel approved the agenda, but modified the order of presentations. THE APPROVED AGENDA REFLECTS THE ORDER LISTED IN THESE MINUTES.
2. Discuss Plant Status and Inspector Insights and Emergent Issues List Mr. Rutkowski briefed the Panel on the plant status and emergent issues. The Panel discussed an inspection plan to provide expanded coverage during the upcoming Christmas weekend. THE APPROVED PLAN IS ATTACHED TO THESE MINUTES.

Mr. Grobe led a discussion of a teleconference conducted earlier today with Mr. Lew Myers. Mr. Myers discussed plans for plant heat up and stated that his staff plans to perform an internal assessment of plant activities through Mode 3 and contract an independent review of that assessment. Mr. Myers stated that he will be prepared to present the results to the NRC during a future public meeting.

3. Discuss New/Potential Licensing Issues - Update on SG Inspection RAI status Mr. Hopkins led a discussion of new/potential licensing issues, including an update on a Request for Additional Information regarding steam generator inspections. Mr. Hopkins stated that the licensee replied to six questions related to restart and the NRC accepted the licensee's responses. Mr. Hopkins also stated that there are an additional eight questions outstanding which the licensee is expected to submit for NRC review by the end of December 2003.
4. Discuss Communication Status - Update on Status of Letter submitted by Ohio Citizen Action Group Memo The Panel discussed the following memorandum:

TO: James Caldwell, Administrator, Region III U.S. Nuclear Regulatory Commission

CC: James Dyer, Director, Office of Nuclear Reactor Regulation Samuel Collins, Deputy Executive Director for Reactor Programs Members, Davis-Besse 0350 Oversight Panel FR: Shari Weir, Cleveland Program Director Paul Ryder, Communication Director Ohio Citizen Action DT: December 15, 2003 RE: An analysis of FirstEnergys Davis-Besse 'safety culture' surveys:

- Company figures show continuing retaliation against plant employees

- Data cooked to improve results: Raw data show no improvement The Panel determined that the memorandum did not contain allegations. Mr. Passehl took action to have Pat Buckley enter this memorandum into the action item tracking system for response, due prior to restart (due date of January 19, 2004 for planning purposes).

The Panel also discussed a December 18, 2003, letter from the Union of Concerned Scientists entitled: "Apparent Flaw in Davis-Besse Root Cause Analysis." Mr. Grobe discussed this letter with Mr. Brent Clayton. Mr. Grobe stated that Mr. Clayton's threshold for whether this letter contains allegations hinges on whether the letter contains new information. If all of these issues described in the UCS letter are issues the Panel already considered in evaluating the technical root cause, then the issues may not be allegations. However, the letter still may warrant a response from the NRC.

Mr. Grobe stated that the determination of whether there are any new issues should be handled by NRR. The Panel still needs to consider the issues within the context of late breaking issues regardless of the allegation determination and make a decision on what needs to be done before restart. Mr. Mendiola took action to discuss how the EDO's office plans on handling the letter.

5. Discuss Items for RAM Closure Mr. Phillips led a discussion of RAM Closure Forms. RAM CLOSURE FORMS APPROVED BY THE PANEL ARE ATTACHED TO THESE MINUTES.
6. Update Panel on HPI Design Mod./Post Mod. Acceptance Testing Adequacy Mr. Mendiola led a discussion on the status of the HPI design modification and post-modification acceptance testing adequacy. Mr. Hopkins took action to set up a discussion on January 5, 2004, regarding the NRC review of the engineering design modification package and the NRC review of the minimum recirculation flow analysis for the HPI pumps. The Panel determined that licensee participants in the meeting should include: L. Myers, J. Powers, S. Loehlein, J. Hagan, and J.D. Wilcox.
7. Update Panel on HPI Minimum Recirc Flow Results Mrs. Lougheed updated the Panel on HPI pump minimum recirculation flow results.

Mrs. Lougheed stated that she does not have any operability concerns with HPI pump minimum flow recirculation. The licensee's contractor who implemented the design improvements to the HPI pumps intends to provide to Mrs. Lougheed a written basis for HPI pump operability.

8. Brief Panel on Meeting with S. Buchanan of Ohio Citizens Action Mr. Grobe briefed the Panel on his and Mr. Caldwells meeting with Ms. Sandy Buchanan of Ohio Citizens Action. Ms. Buchanan met with J. Grobe and J. Caldwell on December 23, 2003, to discuss her groups concerns regarding restart.

Mr. Grobe mentioned that Ms. Buchanan is owed a telephone call to inform her when the restart meeting is scheduled.

9. Panel Discussion of Follow up to Restart Readiness Assessment Inspection and Management and Human Performance Inspection Mr. Grobe led a discussion of the plan for performing followup assessments to the restart readiness assessment inspection and management and human performance inspection. The Panel discussed conducting these followup inspections during the week of January 12, 2004.
10. Discuss Action Items The Panel reviewed the following open Action Items with comments as noted:

Item 217 (Closed) - Review and document the acceptability of the licensees withdrawal of the single safety group of control rods to provide a prompt trip response source of negative reactivity. The review will be documented in a resident inspection report.

(10/09)

Addressed in NRC Inspection Report 05000346/2003-022.

Item 221 (Closed) - Research use of a Quick Look letter which formalizes preliminary inspection results prior to final report being issued to address urgent Restart decision issues. (10/28)

Mr. Passehl received copies of letters from S. Texas and discussed with Panel members.

Item 224 (Open) - Rewrite the proposed IN on TSP to be generic and reflect attainable plant conditions and what information should be disseminated to the industry concerning Boric Acid Corrosion Control Programs. (12/09)

This issue is to be discussed internally in Region III on 01/06/04 with a final decision on how to proceed.

Item 227 (Closed) - Leads take actions to bring open Restart Checklist items to Panel, to include actions to resolve closure and people to brief Panel based on leave schedules. (12/16)

Item 230 (Closed) - Develop Action Plan for documenting basis for path forward with licensee based on current inspection results and readiness for restart, to include process for informing licensee. (12/17)

Closed based on todays teleconference with L. Myers and as documented in 12/24/03 Weekly Status Report.

11. Discuss/Update Milestones and Commitments The Panel reviewed and discussed upcoming milestones and commitments.

INSPECTION PLAN HEAT-UP AND MODE CHANGE INSPECTION Davis-Besse Nuclear Power Station Inspection Objectives The objective is to observe licensee performance during the transition to Modes 4 and 3 and subsequent plant response. The inspectors will focus on control room observations of the operating crews, including turnover and/or scheduling meetings, pre-job briefs, operator performance, plant lineups, and control rod drive insertion time testing. Observations will not be limited to the control room and will include tours of the auxiliary building and containment and attendance at scheduled planning and trouble shooting meetings.

Inspection Dates:

The inspection starts approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to Mode 4 (approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on December 26, 2003 based on the licensees current schedule) and continues through NOP/NOT (Sunday, December 28 based on the licensees current schedule). If the schedule is delayed the implementation of this plan would be delayed to coincide with the licensees revised schedule.

Licensees Schedule:

The licensee's latest schedule shows entering Mode 4 during the evening of Friday, December 26, at approximately 2100 local time. Mode 3 is scheduled for the morning of Saturday, December 27. NOP/NOT is scheduled for the evening of Sunday, December 28.

Control Rod Drive Insertion Time Testing is scheduled for mid-week next week and coverage would be separately scheduled for this activity. The plant is currently experiencing problems with its auxiliary boiler which may require shutting down the boiler and breaking condenser vacuum which would impact the presently defined heat-up schedule.

NRC Inspectors and Assigned Shifts:

Should critical activities occur between the hours of 10:00 p.m. and 4:00 a.m., the team will adjust work hours to observe the activities.

Shift Coverage Stephen Campbell, Fermi SRI 4 a.m. to 1 p.m. 419-537-9569 (home) 419-297-9057 (cell)

Jack Rutkowski, Davis-Besse RI 1 p.m. to 10 p. m. 419-693-7571 (home) 614-537-4461 (cell)

Augmented Day Coverage Monica Salter-Williams, Davis-Besse RI 7 a.m. to 5 p. m. 484-645-4206 (cell) 484-645-4207 (cell)

Reviewed: _/RA by Dave Passhel Acting for Christine Lipa/ 12/23/03 Chief, Reactor Projects Branch 4 Approved: __/RA/______________________ 12/23/03 Chairman, Davis-Besse Oversight Panel

December 23, 2003 DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 197 Develop a J. Stang 6/24-Lead changed; 08/21-Lead communication plan changed; 09/30-Discussed, list of Q &

with restart Qs and As. As is being gathered for review and (06/17) forwarding to RA; 10/14-Discussed, J.

Shea is compiling the list of Q & As for review by the Panel and results will be forwarded to the RA; 10/21-Brainstorming session to occur 10/23 to final presentation to Panel; 11/20-The list of Q&As will be inserted to the Comm Matrix today; 12/15-Discussed, Plan with Panel Chairman.

208 Evaluate the need to M. Phillips 10/14-Investigation is still ongoing; call back CI regarding 12/23-Discussed, awaiting DOJ Allegation RIII-2002-A- Investigation.

0177 (D-B) after the OI Investigation is complete (08/21) 212 Determine whether the J. Stang 10/14-Discussed, Set of bullets still Communication Team under development; item will be has received all discussed at next Panel meeting on electronic and written 10/16; 11/04 -Discussed, J. Stang is correspondence from adding to Comm. Matrix; 11/20-Only external sources. If remaining is the documented criteria there is reasonable for proof of reasonable confidence; confidence that the 12/15-Discussed, all inputs received Communication Team from panel members-closure memo has all the to document completeness correspondence then confidence in draft and will go to develop a set of bullets Panel next week.

explaining why there is reasonable confidence.

(09/23)

December 23, 2003 DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 217 Review and document S. Thomas 10/14-Discussed, This review is the acceptability of the ongoing and will be documented in licensees withdrawal of Inspection Report 03-22; 12/15-the single safety group Discussed, issue address in IR 03-22 of control rods to which will be issued next week-brief provide a prompt trip and closure next week; 12/23-response source of Discussed, addressed in NRC negative reactivity. The IR05000346/2003-022. Panel review will be decided this item is Closed.

documented in a resident inspection report. (10/09) 219 Brief Jim Caldwell on S. Thomas 10/28-Brief will include research how Immediate Action information on Exelon approach; Maintenance issue was 11/20-NRC is reviewing a copy of the resolved. He would like licensees revised procedure; 12/15-to see the revised Discussed, FENOC rep met with SRI procedure. (10/21) 12/14 to review new procedure; 12/18-Discussed, need to setup briefing for RA.

220 Develop inspection plan D. Passehl 11/20-The plan is being developed requirements which and supplemented from baseline include review of post requirements; 12/15-Discussed, draft restart security program plan in final.

effectiveness. (10/28) 221 Research use of a D. Passehl 11/20-Awaiting information from STP, Quick Look letter and Millstone restart documents-will which formalizes update the Panel at 11-25-03 preliminary inspection meeting; 12/15-Discussed, working; results prior to final 12/23-Discussed, D. Passehl report being issued to received letters from STP and briefed address urgent Restart Panel. Panel decided this item is decision issues. (10/28) Closed.

December 23, 2003 DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 224 Rewrite the proposed D. Hills 12/15-Discussed, D. Hills is working; IN on TSP to be generic 12/23-Discussed, this issue will be and reflect attainable discussed internally in RIII and plant conditions and brought to Panel on 1/06/03 for a final what information should decision on how to proceed.

be disseminated to the industry concerning Boric Acid Corrosion Control Programs (12/09) 227 Leads take actions to See 12/23-Discussed, Punchlist is bring open Restart Punchlist updated and depicts leads for Checklist items to checklist items for closure. Panel Panel, to include decided this item is Closed.

actions to resolve closure and people to brief Panel based on leave schedules.

(12/16) 228 Place all Email requests M. Mitchell sent throughout Agency, responses received, and issue resolutions in ADAMS package for documentation. (12/16) 230 Develop Action Plan for J. Grobe 12/23-Discussed, based on documenting basis for teleconference with L. Myers today path forward with and as documented in the 12/24 licensee based on Weekly Status Report, the Panel current inspection decided this item is Closed.

results and readiness for restart, to include process for informing licensee. (12/17) 231 As soon as the final D. Passehl RRATI report is issued, Email a copy to Rick Jacobs at INPO

RAM Item No. - E-41 Closed: Y Date of E-Mail - 10/06/03 Author - Blanch Description of Issue - Respond to Blanchs questions concerning the reasoning for pulling control rods during cooldown and whether such complies with the licensees TS.

Restart Checklist Item: N/A Description of Resolution - Davis-Besse Operations Procedure DB-OP-06903, Plant Shutdown and Cooldown, Revision 11 provided an option to cooldown the reactor coolant system with the Group 1 Safety Control Rods fully withdrawn. The reason given in the procedure is to provide trippable reactivity prior to the addition of positive reactivity [potentially caused by a boron dilution accident].

The Technical Specifications do not directly address this issue (whether or not the Group 1 Safety Control Rods can be withdrawn during reactor coolant system cooldown). How it does deal with the issue of reactivity control is primarily via Technical Specification 3.1.1.1. This Technical Specification requires for operational Modes 1, 2, 3, 4, and 5 that shutdown margin shall be greater than or equal to 1% k/k. Additionally, by definition, during operational Modes, 3, 4 and 5, Keff is required to be maintained less than 0.99 [Keff = 0.99 is approximately 1%

k/k].

During the transition from operation Mode 3 [Hot Standby] to Mode 5 [Cold Shutdown], even with the Group 1 Safety Rods withdrawn, the reactor coolant system boron concentration provided a reactivity margin to criticality of approximately 7.5% k/k [or Keff of approximately 0.93 at 532 F] and 7.0% k/k [or Keff of approximately 0.935 at 70 F]. This same reactor coolant boron concentration provided an approximate 2.0% k/k [or Keff of approximately 0.98]

with all the control rods fully withdrawn.

The advantages of this practice were unclear and as a result, the licensee modified their procedures to prevent any Safety Control Rods being withdrawn during plant cooldown. This issue was documented in Inspection Report no-346/2003-022, a copy of which will be provided to Mr. Blanch as he is on distribution for that report.

Reference Material - NRC Inspection Report No. 50-346/2003-022.

RAM Item No. - C-03 Closed: Y Description of Issue - Self-assessment/ISEG - why not effective: Look at commitments made to justify removal of ISEG from TS and whether they are still meeting those commitments.

Description of Resolution - Of note, the ISEG was eliminated as a requirement for Davis-Besse several years ago. The description into why the licensees past oversight activities were not effective is contained in the licensees root cause analyses reports. The NRC has reviewed the reports and determined that they appropriately identified the root causes behind why the

head degradation event occurred. Documentation of the NRCs review of the root cause reports is contained in the Management and Human Performance Phase I and II inspection reports (02-15 & 02-18). To provide the equivalent form of review that was in the past required of the ISEG, the licensee has put in place several organizational review processes. These include the utilization of an Engineering Assessment Board, the Corrective Action Review Board, and several management changes to provide specific oversight for self-assessment activities. The adequacy of the licensees efforts in correcting the root and contributing causes that self-assessment played in the reactor head degradation event was evaluated to address restart checklist item 3.c., Self-Assessment Programs. The inspection results concluded that the licensees corrective actions to improve its self-assessment programs were adequate.

Reference Material - NRC Inspection Report Nos. 50-346/2002-015 (ADAMS Accession No.

ml0330380037), 50-346/2002-018 (ADAMS Accession No. ml032050528), and 50-346/2003-023 (ADAMS Accession No. ml033421074).

RAM Item No. - C-26 Closed: Y Description of Issue - Electrical Distribution System Analysis (coordination, load flow, degraded voltage, fault protection, and ampacity). Hardware fixes may be required based on calc. results.

Description of Resolution - The Electrical Issues Inspection inspected the issues contained in this RAM item. On site, the team reviewed the following calculations:

1. Revision 0 of the AC distribution system calculations. This calculation was performed using ETAP. (The following calculations were not performed using ETAP: cable ampacity, EDG transient analysis, 120 VAC distribution, protective relaying and coordination, bus fast transfers.) Revision 1 of the ETAP calculation was also subsequently reviewed during the later portions of the CATI.
2. Revision 0 of the EDG steady state analysis.
3. Revision 0 of the EDG transient analysis, performed by MPR not using ETAP While a review of the cable ampacity calculations (hand calculations) was not performed, a review of the methodology for the calculation of cable ampacities was performed.

The following calculations were not reviewed:

1. Cable ampacities.
2. Protective relaying and coordination.
3. 120 VAC distribution.
4. DC distribution system calculations.
5. Most recent ETAP calculation (Revisions 2 and 3).
6. EDG Backfeed Loading (Appendix R worst case loading).

The team found that the ETAP calculations generally were adequately performed and that the calculations assumptions were generally well supported and documented. The short circuit and degraded voltage analyses results appeared adequate. These calculations concluded that

transformers were not overloaded and that short circuit ratings for equipment important to safety were not exceeded.

As documented in condition reports, the licensees ETAP analyses determined that if an SFAS level 4 actuation occurred when the plant was operating in Mode 1 at 100 percent power with degraded grid voltage (98.3 percent of nominal) and the electrical distribution systems aligned to a single startup transformer, the resultant reduction in voltage at the essential 4160 V buses would be of such magnitude and duration that the undervoltage relays would have tripped the supply breakers to the essential 4160 V buses. This would have resulted in an unanticipated Loss of Offsite Power. Additionally, if an SFAS level 4 actuation occurred when the plant was operating in Mode 1 at 100 percent power with degraded grid voltage (98.3 percent of nominal) and both startup transformers available, grid voltage would recover; however, voltages on the 480 V essential buses may not recover to the level necessary for the satisfactory operation of some essential loads. These conditions were reported in LER 2003-007 and are being resolved under RAM Item LER-16.

The steady state analysis of the EDGs determined that the loadings during the transient operation of the EDG under LOOP/LOCA conditions were acceptable for proper operation of all safety loads.

The transient analysis of the EDGs determined that the voltage and frequency excursions during the transient loading of the EDG under LOOP/LOCA conditions were acceptable for proper operation of all safety loads. In addition, with the exception of the first load step, all frequency and voltage transient deviations from nominal were in accordance with the requirements of NRC Safety Guide 9 (frequency dip to less than 95% and voltage dip to less than 75%). This frequency and voltage dip is being resolved under RAM Item C-32.

Inspection of the AC Distribution System produced no findings during the Electrical Issues Inspection. Based upon predominantly satisfactory results from the electrical issues inspection, and based upon the licensees corrective action plan for the electrical distribution system appearing reasonable and adequate, this item is closed.

Reference Material - CR 02-05385.

RAM Item No. - LER-13 Closed: Y Description of Issue - Temperature elements TERC3A5 and TERC3A6 [reactor coolant loop 2 hot leg wide range temperature elements], had not had their calibration verified as required by Technical Specification 3.3.3.6. This was due to these two RTDs not being included in the procedure that performs the Technical Specification required calibration and stability checks for reactor coolant system RTDs.

Description of Resolution - Upon discovery of this issue, the licensee verified the calibration of TERC3A5. Calibration verification was not possible for TERC3A6 due to the RTD being damaged during removal. Based on operational data obtained from all four reactor coolant hot leg wide range temperature instruments, the licensee believes that TERC3A6 would have passed its calibration verification. Since TERC3A6 was not calibrated in accordance with Technical Specification surveillance requirements, the licensee determined that the plant had

operated with the non-calibrated instruments which represented a condition prohibited by Technical Specification and reportable under 10 CFR 50.73(a)(2)(i)(B)

The licensee performed an extent of condition and did not identify additional omission of RTD calibration requirements. On May 22, 2003, the licensee implemented procedure DB-SC-03159, RTD Cross Calibration. This procedure replaced the original calibration procedure and will be used to determine the calibration accuracy and stability of the RCS narrow and wide range RTDs. This new procedure included TERC3A5 and TERC3A6 as part of the cross calibration process for reactor coolant system RTDs.

Although the failure to perform the required surveillance testing for TERC3A5 and TERC3A6 was a minor violation of Technical Specification 3.3.3.6, it constituted a violation of minor significance that is not subject to enforcement action in accordance with Section IV of the NRCs Enforcement Policy. The LER was reviewed by the inspectors and no findings of significance were identified.

Reference Material - NRC Inspection Report No. 50-346/2003-025, scheduled to be issued in January 2004.

RAM Item No. - LER-16 Closed: Y Description of Issue - AC System Analysis Results Show Potential Loss of offsite Power Following Design Basis Accident.

Description of Resolution - Two issues were identified in the Licensee Event Report as follows:

  • LER 2003-007-00 reported that if an SFAS level 4 actuation occurred when the plant was operating in Mode 1 at 100 percent power with degraded grid voltage (98.3 percent of nominal) and the electrical distribution systems aligned to a single startup transformer, the resultant reduction in voltage at the essential 4160 V buses would be of such magnitude and duration that the undervoltage relays would have tripped the supply breakers to the essential 4160 V buses.

This would have resulted in an unanticipated Loss of Offsite Power.

  • LER 2003-007-00 reported that if an SFAS level 4 actuation occurred when the plant was operating in Mode 1 at 100 percent power with degraded grid voltage (98.3 percent of nominal) and both startup transformers available, grid voltage would recover; however, voltages on the 480 V essential buses may not recover to the level necessary for the satisfactory operation of some essential loads.

For the first issue, the licensee discovered that the plant operated with only one startup transformer available several times in the past. Specifically, this condition existed on August 7, 2001, January 5, 2002, and August 13 and 14, 2001. Because of the previously described condition, neither offsite source was operable during these periods.

Additionally, during the August 13 and 14 time period, since this condition was not known at the time, the EDGs were not tested within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, nor was the plant shut

down to hot standby (Mode 3) conditions, as required by LCO 3.8.1 for a loss of both offsite sources. This constituted a violation of Technical Specifications.

For the second issue, the licensee, identified that under degraded voltage conditions, the following equipment may not be available: the control circuitry for the Main Feedwater Steam Generator Isolation Valves, one train of the Component Cooling Water Ventilation System, and one train of the Emergency Ventilation System. Since only one train of each of the ventilation systems may have been disabled, the remaining trains would be available to perform the intended function.

The inspector reviewed this LER and determined that the corrective actions associated with the issues reported appeared to be reasonable and adequate. The following corrective actions had already been completed:

  • Change the tap settings on the 4160 V to 480 V essential substation transformers supplying power to essential buses E1 and F1 to increase the 480 V essential bus voltage during power operations.
  • Change the tap settings on the 480 V to 240 V transformers feeding essential buses YE2 and YF2 to increase the 240 V essential bus voltage.
  • Install shorting bars for selected hydramotor circuits.
  • Revise the trip setpoint and Allowable Value for the 90% undervoltage essential bus feeder trip relays.
  • Until the change to the Technical Specifications for the 90% undervoltage essential bus feeder trip relays is implemented, maintain administrative controls in accordance with NRC administrative Letter 98-10.

Additionally, the closure packages for these corrective actions associated with the LER were reviewed.

In their Integrated Report to Support Restart dated November 23, 2003, the licensee has committed to submit a license amendment request for revisions to the SFAS Technical Specification values by January 30, 2004.

Reference Material: - None.

RAM Item No. - NCV-29 Closed: Y Description of Issue - No Procedural Guidance for Performing Immediate Action Maintenance.

Description of Resolution - The performance deficiency associated with this issue was that the licensee failed to provide procedural guidance on how maintenance performed utilizing the Immediate Action Maintenance process was controlled, reviewed, or tested, to verify the adequacy of the maintenance activity. The inspectors reviewed a new licensee procedure NOP-WM-4002, Repair Identification and Toolpouch Maintenance, Revision 2, and found the procedure sufficiently outlined the process to be used when the Immediate Action Maintenance process was implemented.

Reference Material - NRC Inspection Report No. 50-346/2003-018 (ADAMS Accession No.

ml033080433) and licensee condition report nos. 03-08776, 03-08622, and 03-08791 RAM Item No. - NCV-30 Closed: Y Description of Issue - Improper Implementation of DB-OP-00000, Conduct of Operations, Revision 06, pertaining to the implementation of the Immediate Action Maintenance (IAM) process. This process was improperly implemented to perform routine adjustments on an auxiliary feedwater pump governor.

Description of Resolution - The performance deficiency associated with this event is the senior operations management inappropriately authorized the performance of the Immediate Action Maintenance process to perform adjustments on 1 turbine driven auxiliary feedwater pump governor. The inspectors reviewed procedure DB-OP-00000, Conduct of Operations, Revision 07. This revision eliminated the improper entry criteria for the use of the Immediate Action Maintenance Process. The new revision specifically states that work performed will categorized as Priority 100 - Immediate. Procedure NOP-WM-4002, Repair Identification and Toolpouch Maintenance, Revision 01, defined Priority 100 work as A condition which is an immediate or imminent threat to nuclear safety or personnel/public safety. Work the necessary resources 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day to achieve completion at the earliest possible time.

Reference Material - NRC Inspection Report No. 50-346/2003-018 (ADAMS Accession No.

ml033080433) and licensee condition reports (CRs) 03-08776, 03-08622, and 03-08791.

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