ML040760266

Revised Minutes of Internal Meeting of the Davis-Besse Oversight Panel (Revised RAM Closure Attachment)

ML040760266
Person / Time
Site:	Davis Besse
Issue date:	03/11/2004
From:	Grobe J NRC/RGN-III
To:	NRC/RGN-III
References
Download: ML040760266 (25)

Text

March 11, 2004 MEMORANDUM TO: Davis-Besse Nuclear Power Station IMC 0350 Panel FROM: John A. Grobe, Chairman, Davis-Besse Oversight Panel /RA C.

Lipa Acting for/

SUBJECT:

REVISED MAY 27, 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 May 27, 2003. Attached for your information are the minutes from the internal meeting of the Davis-Besse Oversight Panel, the Restart Action Matrix (RAM) Items approved for closure, and the Open Action Items List.

Attachment:

As stated cc w/att: H. Nieh, OEDO J. Dyer, RIII J. Caldwell, RIII R. Gardner, DRS B. Clayton, EICS G. Wright, DRP DB0350

MEMORANDUM TO: Davis-Besse Nuclear Power Station IMC 0350 Panel FROM: John A. Grobe, Chairman, Davis-Besse Oversight Panel

SUBJECT:

REVISED MAY 27, 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 May 27, 2003. Attached for your information are the minutes from the internal meeting of the Davis-Besse Oversight Panel, the Restart Action Matrix (RAM) Items approved for closure, and the Open Action Items List.

Attachment:

As stated cc w/att: J. Dyer, RIII J. Caldwell, RIII R. Gardner, DRS B. Clayton, EICS G. Wright, DRP DB0350 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML040760266.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/ske CLipa JGrobe /RA C. Lipa Acting for/

DATE 03/11 /04 03/11 /04 03/11 /04 OFFICIAL RECORD COPY

MEETING MINUTES: Internal IMC 0350 Oversight Panel Meeting Davis-Besse Nuclear Power Station DATE: May 27, 2003 TIME: 1:00 p.m. Central ATTENDEES:

J. Lara M. Phillips C. Lipa D. Passehl W. Ruland T. Mendiola S. Thomas B. Daly J. Hopkins Agenda Items:

1. Discuss Closed Restart Action Matrix (RAM) Items and Closure Forms M. Phillips presented and the Panel discussed a list of RAM items for closure. THE RAM ITEMS APPROVED FOR CLOSURE ARE ATTACHED TO THESE MINUTES.

2. Discuss New Allegations The Panel discussed no new allegations.

3. Discuss Any Allegations for Which an Extension Was Requested M. Phillips presented this discussion. There were no allegations for which an extension was requested.

4. Discuss Plant Status and Inspector Insights and Emergent Issues List S. Thomas discussed plant status and inspector insights and emergent issues.

Messrs. Lara and Daly discussed a potential issue pertaining to 10 CFR 50 Appendix R Section III-L, "Alternative and Dedicated Shutdown Capability." A specific concern was identified for the shutdown function performance goal of maintaining reactor coolant level. An NRC safety evaluation report issued in 1991 apparently allows the licensee to maintain reactor coolant level above the top of active fuel instead of maintaining level within the range of indication in the pressurizer.

NEW ACTION ITEM - T. Mendiola to brief Panel during its next meeting on need for a TIA on a potential issue pertaining to 10 CFR 50 Appendix R Section III-L, "Alternative and Dedicated Shutdown Capability." A specific concern was identified for the shutdown function performance goal of maintaining reactor coolant level. An NRC safety evaluation report issued in 1991 apparently allows the licensee to maintain reactor coolant level above the top of active fuel instead of maintaining level within the range of indication in the pressurizer.

5. Discuss Action Items The Panel discussed the list of action items and their status.

Item 97 - Open. NRR indicated that response and acceptance of the bulletins is a compilation effort rather than specific review of licensee responses. Panel determined that to close out need acknowledgment letter that the bulletin responses have been received.

Item 136 - Reopen. - J. Grobe to send an email week of 05/19 with Panels question of whether the licensees method and criteria are adequate to address the TS for zero unidentified reactor coolant leakage from the reactor coolant system. By the end of the week of 05/19 NRR to formulate a proposed position for discussion. The position by NRR will be discussed by the Panel during the 05/27 0350-Panel meeting. After discussing on 05/27 the issue is to be presented internally within the NRC for dissenting views.

Item 145 - Open. Discuss during Panel Meeting on June 10.

Item 189 - Open. Bill Ruland to discuss with Dave Beaulieau.

Item 190 - Close. Action Complete.

6. Discuss New/Potential Licensing Issues J. Hopkins discussed the status of licensing issues.

7. Discuss Items for Licensee Weekly Calls The Panel discussed several items for the licensee weekly calls.

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

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 24a Discuss making Panel Discuss by June 30, after safety information related to significance assessment complete; HQ/licensee calls publicly 6/27 - Invite Bateman to panel mtg.

available To discuss what else is needed to closeout the CAL (i.e. quarantine plan); 7/2 - NRR not yet ready to discuss; 7/16 - See if procedures have changed on CAL closeout -

does JD need to send letter?; 7/18

- Discussed - is there an applicable regional procedure?; 8/6 -

Discussed. Need to determine the final approach on the core removed from the head and the final approach on the head before the quarantine can be lifted; 8/22 -

Revisit action item after letter sent to licensee confirming plans with old vessel head (head may be onsite longer than originally anticipated); 8/29 - Memo to be sent to Region, with a letter to go out next week; 10/01- Discussed.

1) Conduct NRC staff survey-due 10/7 2)Memo to NRR - due 10/11

3) Region to issue letter; 11/07-Letter required from NRR on head quarantine status; 11/19 - Letter in draft; 01/03 - A. Mendiola to look at phone conference writeups on quarantine decision making to determine if they can be released to the public; 01/07 - discussed; 01/21 - discussed; 01/31- A.

Mendiolas action; 02/11 -

Completion of Licensee Phase 3 sampling plan required; 02/21 -

17.5 Rem to cut samples, Less samples may be required; 04/03 -

Completion of Phase 3 sampling plan scheduled for late April -

discuss again then; 04/08 - Revisit in June 2003

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 54a Review TSP amendment D. Pickett 7/9 - Discussed. Will wait for and advise the panel on response from licensee; 7/16 -

the need for a TIA on Discussed - added action item 54b; Davis-Besse (7/2) 8/6 - Sent to the licensee on 7/22 and a response is due by 8/22; 8/22 - Discussed - need to check if response has been received; 8/27

- Received response - DRS is reviewing - will fax to NRR for 54b; 8/29 - Discussed, DRS report of response to be issued to panel prior to item 54b; 10/1-Discussed.

DRS coordinating with NRR 11/07-Discussed - On hold for draft with specific information; 12/10 - B.

Dean believed B. Bateman thought a calculation for sufficient volume of TSP was completed to technical specification value. However questions whether the calculation was to technical specification or actual TSP level remain; 01/03 -

Item under NRR review.

Calculation completion expected on Jan 17. Allegation issue in RIII domain; 01/07 - Allegation Item #3 under NRR Review for Resolution; 01/21 - Item #3 is under Region III control for final letter, holding for NRR input; 02/11 - Writeup for NRR input provided 4 answers, going back to reviewer to ensure specific tasking is clear to answer allegation concerns. Action item 54c created; 02/21 - Allegation at 242 day mark. Effective expression of due date required; 04/03 -

Messrs. Beckner, Grobe, Ruland,

& Mendiola to discuss; 04/08 - J.

Hopkins to set up call for 4/11/03

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 73 Send feedback form on Lipa 8/6 - Generate feedback after IMC 0350 procedure to Mendiola panel meetings reduced to once IIPB (8/6) per week; 8/29 - Discussed - no change; 10/1 - Discussed; 11/7 -

D Passehl sent email to C Carpenter and D Coe indicating that we would be able to perform a review of the draft IMC 0350 during the first quarter of 2003; 12/3- discussed; 01/03 - 2 parts, short part- C. Lipa with P. Harris, long part- B. Dean; 01/07 - 2nd larger response will require meeting between all parties; 01/21

- Communications with P. Harris; 01/31-Meeting with P. Harris on Feb 4; 02/11 - Many concerns identified by the panel for inclusion; 02/21 - July 1 due date for larger input.

97 Bulletins 2002-01 and NRR 11/07 - Discussed, further 2002-02 response and research and discussion required; acceptance (9/5) 01/07 - RAI response expected Mid February; 01/31- On track; 02/11 - New Orders will supercede BL2002-01 and BL2002-02 responses with the exception of the BL2002-01 Boric Acid Corrosion program information request; 02/21 - Licensee RAI response delayed. Both Order and BL2002-01 Boric Acid Corrosion program responses to be tracked as RAM items; 04/08 - Discussed, leave open pending NRR review; 05/27 - Discussed - close out need acknowledgment letter that the bulletin responses have been received 126 Review Davis- Strasma Besse/Vessel Head 02/11 - Checked, but revisiting Degradation web site item; 02/21 - Web site being content for ease of use by reassessed.

the public. (11/07)

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 136 NRR acceptance of NOP W. Dean/J. 01/07 - Item discussed. Meeting criteria and method Grobe summary of November 26, 2002 (01/03) meeting has notation of NRR staff J. Grobe to send an email impressions of test plan. Once week of 05/19 with drafted, issue will be surveyed to question of whether the staff to determine if consensus is licensees method and correct; 01/21 - Meeting summary criteria are adequate to to discuss Flus System, Test address the TS for zero agreement, and future inspections; unidentified reactor 1/31 - T. Chan fwd to J. Hopkins; coolant leakage from the 2/11 - J. Jacobson questions need reactor coolant system. to be folded in (chem-wipes); 2/21 By the end of the week of - Polling of staff discussed; 2/24 -

05/19 NRR to formulate a Polling of staff by March 7; 3/25 -

proposed position for Staff to be polled after 4/4/03 discussion. The position meeting in headquarters, and by NRR will be discussed meeting should address whether a by the Panel during the rational basis exists that the 05/27 0350-Panel bottom head is not leaking, and meeting. After discussing whether a critical flaw size will not on 05/27 the issue is to be appear during the next operating presented internally within cycle; 04/08 - J. Hopkins writing the NRC for dissenting mtg summary for 4/4, licensee to views address additional questions on (05/27 - Reopen) 4/9 telephone call; 05/16 - Closed; 05/27 - Reopen.

138 Evaluate the effectiveness A. Mendiola, 01/31 - Ongoing; 02/21 - New EDO of the Comm Plan (01/07) C. Lipa Comm Plan for Crisis Update, A.

Mendiola to review for inclusion.

145 Prepare a special D. Passehl 02/21 - date to be determined; inspection plan for the 05/27 - DIscussed.

restart readiness team inspection. (01/09)

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 147 Generate a list of items to D. Passehl 01/31 - working; consider after restart as 02/11 - Include dates and well as transition back to deadlines to Manual Chapter 0350 the normal 0350 when restart inspections planner terminating the 0350 Panel. The items should include plans to augment inspection of corrective actions, inservice inspection, and safety culture monitoring.

(01/09) 156 Read Generic Safety J. Hopkins 01/21 - Determine status of GSI-Issue-191, "Assessment of 191; 02/21 - Check GL98-04 Debris Accumulation on response on coatings. Draft GL PWR Sump Pump and Draft Reg Guide needs review Performance" (01/09) for DB relevance; 02/24 - Request Response Review and Program Implementation to GL98-04; 03/04

- activity to be reassigned to Reactor Engineer who will close sump LER; 04/08, D. Hills to discuss with K. Coyne and A.

Dunlop work assignments 174 Review 2/4 transcript for R. Lickus Mr. Witts recommendations (2/18) 178 Determine the type of C. Lipa backlog assessment that will be performed and by whom. Two attributes need to be considered: (1) the capability of the licensee to manage the backlog in an operating environment; and (2) the impact of the backlog on equipment reliability.

(03/04)

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 186 Add Dennis Kucinich to A. Saso the standard distribution list on documents for Davis-Besse. Then remove Dennis Kucinich from distribution 90 days after the NRC reply to his 10 CFR 2.206 Petition is signed out. (04/22) 187 Arrange for Fire Protection R.Gardner Inspection to evaluate the licensees safe shutdown design calculations in the context of electrical and thermo-hydraulic design to support closure of Restart Checklist Item 5.b, "Systems Readiness for Restart." (04/29) 188 Arrange for Radiation K.Riemer Protection Inspection to evaluate stability in Radiation Protection organization effectiveness.

An inspection plan is to be drafted and presented to the Panel for approval, including applicable Restart Action Matrix items, for an inspection to be conducted in the July 2003 time frame. (04/29) 189 Investigate how the NRC B.Ruland 05/27-Discussed handled communication of potential inspection issues and findings during in-progress inspection work at Millstone in regards to their organization and human performance problems. (05/16)

DAVIS-BESSE OVERSIGHT PANEL OPEN ACTION ITEM LIST Item Action Item (Date Assigned to Comments Number generated) 190 Discuss with T. Kozak C.Lipa 05/27-Discussed removing Davis-Besse from the Operational Management Information Report (OMI). (05/16) 191 Prepare Questions and C.Lipa/

Answers (Qs and As) A.Mendiola once the GT 221 letter to Rep. Kucinich is issued to ensure consistent communications regarding how the NRC articulates issues to the public. (05/16) 192 Draft an update to the D. Passehl March 31, 2003, inspection schedule letter.

(05/16) 193 Consider TIA on an issue T. Mendiola A concern was identified for the pertaining to 10 CFR 50 shutdown function performance Appendix R Section III-L, goal of maintaining reactor coolant "Alternative and Dedicated level. An NRC safety evaluation Shutdown Capability." report issued in 1991 apparently (05/27) allows the licensee to maintain reactor coolant level above the top of active fuel instead of maintaining level within the range of indication in the pressurizer.

RAM Item No. - E-30 Closed: Y Date of E-Mail - 05/02/03 Author - Mr. Lochbaum Description of Issue - What is the status of generic communications regarding the containment sump performance.

Restart Checklist Item: N/A Description of Resolution - Issue not a Davis-Besse restart issue. Issue relates to desire on the part of Mr. Lochbaum to keep up with generic issue resolution. Issue is closed for purpose of Davis-Besse restart.

Reference Material - Platts Inside NRC, May 5, 2003 edition, article entitled Forthcoming Containment Sump Blockage Bulletin Surprises Industry RAM Item No. - L-38 Closed: Y Date of Letter - 07/15/02 Author - UCS-15 Description of Issue - Could operators execute emergency procedures with containment radiation monitors having a mean-time-between-failures of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?

Restart Checklist Item: N/A Description of Resolution - There are no emergency procedures which utilize the containment radiation monitors as a primary indication. Rather, the monitors used for post-accident response are direct reading gamma radiation dome monitors, which are not affected by particulates in containment. Further, there are no control board alarms associated with the containment radiation monitors, only a computer alarm. This computer alarm is used for information only. If the alarm actuates, the operators verify no low flow conditions exist, then determine which channel is in alarm. Subsequent actions, which are non-emergency, are determined by which channel (i.e., particulate, iodine or noble gases) is in alarm.

Reference Material - None.

RAM Item No. - L-39 Closed: Y Date of Letter - 07/15/02 Author - UCS-16 Description of Issue - Did operability evaluations performed for the Containment Rad monitors consider monitors ability to operate in the atmospheric conditions following an accident?

Restart Checklist Item: N/A Description of Resolution - Since the containment radiation monitors have never been inoperable except for maintenance, including filter changeouts, no operability evaluations have ever been completed. The licensee monitored flow, and changed the filters before the low flow alarm actuates.

Reference Material - None.

RAM Item No. - L-40 Closed: Y Date of Letter - 07/15/02 Author - UCS-17a Description of Issue - Does FENOC have an engineering calculation/analysis of non-radioactive particulate matter in containment atmosphere following DBAs?

Restart Checklist Item: N/A Description of Resolution - An inspector discussed this issue with engineering personnel and determined that the licensee does not have an engineering analysis for non-radioactive particulate matter in the containment atmosphere following a DBA other than the sump transport analysis. This is because the system will not operate during a LOCA. The SFAS signal would isolate the sample lines, which would have to be opened manually.

Reference Material - None.

RAM Item No. - L-41 Closed: Y Date of Letter - 07/15/02 Author - UCS-17b Description of Issue - If answer to L-40 is no, how is it assured that the plant is operating within design bases with respect to 10 CFR 50, App. A, GDC4?

Restart Checklist Item: N/A Description of Resolution - The Accident Range Gaseous Effluent Monitor draws gases from an isokinetic sample probe on the inlet line of the normal range monitor to the high range unit through the containment inlet line. These gases, drawn by a metal bellows pump, enter a particulate filter collector, and are retained on filter media. When each of the first two cartridge assemblies are filled, and the flow transfers to the third assembly, the operator must replace the two used filter cartridges. On the inlet and outlet lines of each particulate assembly, manual valves are provided to isolate an assembly for cartridge removal and replacement. For a low

flow condition, such as might occur with a high concentration of non-radioactive particulate matter, a computer alarm would signal the operators to select the next filter, which can be done from the control room panel. Additionally, upon a DBA, the Safety Features Actuation System will isolate containment penetrations, including the Containment Radiation Monitoring System.

Therefore, non-radioactive particulate matter within containment would not be introduced into the sample lines. The sample line isolation valves must be manually reopened to re-initiate containment atmospheric sampling, which is the method that would be used for post-accident sampling.

Reference Material - None.

RAM Item No. - L-43 Closed: Y Date of Letter - 07/15/02 Author - UCS-19 Description of Issue - Has the extent-of-condition assessment by FENOC included verification that atmospheric sampling lines leading to the Normal Range particulate Radiation Skid are free from boric acid corrosion?

Restart Checklist Item: N/A Description of Resolution - The extent of condition assessment did include an assessment of the potential for boric acid in the containment radiation monitoring sample lines. The licensees corrective actions to resolve this issue are described in Condition Report CR 03-02451, Boric Acid Concern with CTMT Rad Monitor Sample Lines. The licensee performed an evaluation which concluded that the sample lines did not need to be cleaned out.

Reference Material - Licensee Condition Report CR 03-02451.

RAM Item No. - L-48 Closed: Y Date of Letter - 7/15/02 Author - UCS Description of Issue - What damaged valve RC-262?

Restart Checklist Item: 3.d & 3.f Description of Resolution - Based on an inspectors review of the applicable condition report, the valve was not damaged. The packing leak off tap was not being used and had a plug screwed into it which leaked. This was the leakage source from the valve, and the location of the tap was such that it did not constitute part of the defined Reactor Coolant System (RCS) pressure boundary.

Reference Material - Licensee Condition Report CR 00-1452.

RAM Item No. - L-50 Closed: Y Date of Letter - 7/15/02 Author - UCS Description of Issue - If boric acid was root cause of RC-262 damage (reference RAM Item No. L-48), doesnt back to back damage to RC-2 and RC-262 suggest FENOCs extent of condition and problem resolution processes are flawed?

Restart Checklist Item: 3.d & 3.f Description of Resolution - Since boric acid was not the cause of RC-262 damage, this item is moot. There is no suggestion that FENOCs extent of condition or problem resolution were flawed.

Reference Material - RAM Item No. L-48.

RAM Item No. - L-51 Closed: Y Date of Letter - 7/15/02 Author - UCS Description of Issue - If boric acid was not cause of damage to RC-262, isnt FENOCs ISI, PM, and aging management flawed?

Restart Checklist Item: 3.d & 3.f Description of Resolution - The location of the leak was not at a pressure boundary point, and as such, would not be observed as part of the ISI inspection. Corrective actions for the licensees ISI inspection program were reviewed by the NRCs programs inspection (see inspection report 03-09) and found to be acceptable. The leak was caused by a bad screw connection at the packing leak off tap plug. Attempt to seal weld leaking threads on packing leak off tap was unsuccessful due to access problems. Subsequently, Furmanite leak seal was used. During the next outage valve bonnet was replaced using live load packing.

Reference Material - RAM Item No. L-48.

RAM Item No. - L-58 Closed: Y Date of Letter - 7/15/02

Author - UCS-29a & UCS-29b Restart Checklist Item: N/A Description of Issue - Does the NRC believe that forcing a company to write 8 CRs really indicates a proper threshold for CRs? If yes, please explain.

Description of Resolution - It is a normal practice for licensees to address NRC inspection findings by the creation of condition reports or similar documents to put the findings into the licensees corrective action system. The NRC does not have a set threshold for the number of condition reports that a licensee should write, nor does the NRC force licensees to write condition reports. The licensee writes condition reports to identify safety issues so that the appropriate assessment and corrective action can be accomplished for each issue. One of the concepts that is an important part of the Reactor Oversight Program, is that the licensee has a robust corrective action program that identifies issues, at the appropriate thresholds, and corrects the identified deficiencies in a timely manner, commensurate with plant safety. This is a requirement of Criterion XVI of Appendix B to 10 CFR Part 50, including for those issues identified by NRC inspections. It is very rare that an NRC inspection does not identify issues, some more serious than others. The NRC would expect the licensee to place those issues into its condition reporting system. As to whether 8 condition reports constitutes a proper threshold for CRs, that depends on the type of inspection conducted and whether the licensee also performed an identical inspection of the same areas.

Reference Material - None.

RAM Item No. - L-70 Closed: Y Date of Letter - 09/13/02 Author - Gurdziel # 5 Description of Issue - Does the corrective action process allow closure of an item to a work order?

Restart Checklist Item No.: N/A Description of Resolution - The answer to the question is yes. The corrective action program (CAP) at Davis-Besse is comprised of three levels of condition reports; Significant Conditions Adverse to Quality (SCAQ), Conditions Adverse to Quality (CAQ), and Condition Not Adverse to Quality (NCAQ). The procedure that defines the CAP at Davis-Besse is NOP-LP-2001, Condition Report Process, Revision 04. Step 4.1.10 of this procedure states all CAFs

[corrective action forms] for condition reports categorized as SCAQ and CAQ shall be tracked in the CREST [Condition Report and Status Tracking] database from the initiation until the approved corrective action(s) are implemented, have corrected the deficiency and their implementation is documented in the database. CAFs for condition reports categorized as NCAQ may be closed to other tracking systems as applicable. This would include a work order. Examples of issues that would be categorized at NCAQ include: minor drawing errors, instrument errors that do not render equipment inoperable, minor personnel contamination

incidents, procedural deficiencies in non-safety related and non-quality procedures with no adverse consequences, data entry errors that have no impact, and certain equipment failures involving components that are Non-Q or Non-AQ, including Maintenance Rule non-safety related functional failures.

Reference Material - Licensee procedure NOP-LP-2001, Condition Report Process.

RAM Item No. - L-81 Closed: Y Date of Letter - 12/16/02 Author - Gurdziel # 17 Description of Issue - For the containment elevator safety glasses incident mentioned in Inspection Report 50-346/02-17, should a for cause fitness-for-duty test have been performed?

Restart Checklist Item: N/A Description of Resolution - Inspectors reviewed the licensee procedure NOP-LP-1002, Fitness for Duty Program, which governs the licensee response to fitness-for-duty issues.

The purpose of this procedure is to provide reasonable assurance that personnel perform their duties in a safe, reliable, and trustworthy manner and are not under the influence of legal or illegal substances or mentally impaired from other causes which would adversely hinder their ability to competently perform their duties. Based on a review of the circumstances surrounding the event, it is the inspectors conclusion that no fitness for duty evaluation was warranted for the licensee personnel involved.

Reference Material - Licensee Procedure NOP-LP-1002, Fitness for Duty Program.

RAM Item No. - C-01 Closed: Y Description of Issue - Completeness and accuracy of licensees response to Generic Letter 97-01 Description of Resolution - Based on a request from the Davis-Besse 0350 panel, A. Dunlop reviewed three questions with respect to Davis-Besses response to Generic Letter 97-01, Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations. The following is the results of the review, which was verbally discussed with the 0350 Panel on September 19, 2002.

1. Did the licensees response to the generic letter modify the licensing basis?

The licensees response to Generic Letter 97-01 should be considered part of the licensing basis based on the following.

In their response dated July 28, 1997, the licensee indicated that Babcock & Wilcox Owners Group (B&WOG) had developed an integrated response, which was documented in B&WOG Topical Report, B&WOG integrated response to Generic Letter 97-01: Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations, BAW-2301, dated July 1997. The licensees letter also indicated that Toledo Edison endorses BAW-2301. The Topical Report provided the justification and schedule for an integrated vessel head penetration inspection program for all B&WOG plants. The Topical Report determined that Davis-Besse was not considered at significant risk to require inspections of the reactor vessel nozzles from beneath the head in the near term (1998-2000). The report did state that B&WOG plants continued to comply with 10 CFR 50.55a and meet the intent of Appendix A General Design Criterion 14, based on visual inspections performed at all B&WOG plants during each refueling outage in accordance with utility responses to Generic Letter 88-05 and perform inspections of the required number of control rod housings during each inspection interval per ASME Code requirements. (It was not verified that these ASME Code inspections were conducted or if the licensee was granted relief from the requirement by the NRC.)

After the licensee, through the B&WOG, responded to a Request for Additional Information (RAI), the NRC approved the licensees response to the generic letter in a letter dated November 29, 1999, which stated the integrated program provides an acceptable basis for evaluating VHPs based on the license endorsing the NEI Submittal of December 11, 1998, (integrated response to RAI) and indicated their participation in the NEI/B&WOG integrated program.

NEI 99-04, Guidelines for Managing NRC Commitment Changes, defined Regulatory Commitment as an explicit statement to take a specific action agreed to, or volunteered by, a licensee and submitted in writing on the docket to the NRC. In addition, the guidance states A Regulatory Commitment is an intentional undertaking by a licensee to ... (2) complete a specific action to address an NRC issue or concern (e.g., generic letter, bulletin, order, etc.). In discussions with the Region III counsel, it was concluded that to be a regulatory commitment as stated in the NEI guidance, the word commitment did not have to be used as part of the explicit statement. The NEI guidance was determined to be acceptable by the NRC in SECY-00-0045, dated February 22, 2000. SECY-98-224 and 10 CFR 54.3 define Current Licensing Basis, in part, as including licensees commitments remaining in effect that were made in docketed licensing correspondence such as licensee repossess to NRC Bulletins, generic letters, and enforcement actions.

As such, based on the definitions included in NEI 99-04 for regulatory commitments and 10 CFR 54.3 for Current Licensing Basis, the licensees response to Generic Letter 97-01 meets the conditions to be considered part of the licensing basis.

This regulatory commitment or change to the licensing basis, however, does not appear to require to be included in the FSAR based on the following.

10 CFR 50.71(e) discusses the requirement to update the FSAR. It states that the submittal shall include the effects (Effects of changes includes appropriate revisions of descriptions in the FSAR such that the FSAR (as updated) is complete and accurate) of: ... all analyses of new safety issues performed by or on behalf of the licensee at Commission request. The generic letter requested the licensee to provide the analysis that supports their intended actions. Since the generic letter discussed the recent concerns with the degradation of CRDM nozzles, this would be considered a new safety issue. The generic letter also request the licensee to provide

an analysis to support their intended actions. As such this would appear to meet the requitement that this analysis should have been included in an update to the FSAR. However, NEI 98-03, Guidelines for Updating Final Safety Analysis Reports, states in section A5, Removing Unnecessary Information from Updated FSARs, that licensees may remove from the UFSAR commitments that are not integral to required UFSAR information, i.e., design bases. SECY-97-036, dated May 20, 1997, supports this by stating the staff should formulate an approach that would ... allow obsolete or less meaningful information and commitments to be readily removed from the FSAR. Since the analysis performed in response to the generic letter was in support of an inspection schedule for the CRDMs and not in itself part of the plants design bases, the guidance would not require its inclusion into the FSAR.

2. If so, was it recognized?

No. Based on statements by the licensee during the August 15, 2002, meeting, the licensee stated that they did not recognize that their response to Generic Letter 97-01 and also Generic Letter 88-05 on the Boric Acid Control Program, were licensing commitments. As such, the licensee did not recognize the commitments made to these generic letters were regulatory commitments and did not control them as such.

3. Were there false statements (in the August 15, 2002, meeting transcript concerning the generic letter)?

A review of the transcripts was performed to determine if any possible false statements were made with respect to the licensees commitment or lack there of to the generic letter. As discussed in item two, the licensee admitted that they did not recognize that their response to Generic Letter 97-01 and also Generic Letter 88-05 on the Boric Acid Control Program, were licensing commitments. Based on a review of the transcript looking for references to the generic letter, there did not appear to be any false statements.

Reference Material - None.

RAM Item No. - C-13 Closed: Y Description of Issue - Process Pressure and Temperature Curves for the "New" Vessel Head.

Description of Resolution - By letter dated January 22, 2003 (serial no. 1-1285), the licensee submitted verification that the pressure/temperature curves in the TS are applicable to the new RPV head. The NRC staff has reviewed the submittal and has determined that the information provided is satisfactory. This conclusion is documented in Inspection Report 03-04.

Reference Material - Inspection Report 03-04, which is in ADAMS as Accession Number ml031320705.

RAM Item No. - C-22 Closed: Y Description of Issue - Pressurizer safety issue - Safety Evaluation for Temporary Modification 98-0036 did not fully address Updated Safety Analysis Report. (LLTF Issue)

Description of Resolution - The Safety Evaluation for TM 98-0036 (SE 98-0045) does adequately address the impact of the temporary modification on the drain line that runs from the 8" pressurizer relief header to the pressurizer quench tank. SE 98-0045 references USAR Section 5.2.4.7 and addresses the impact of this temporary modification. Specifically:

 Section 4.0 - Functions Important to Safety of Affected SSCs: in part, The function that is not important to safety that is affected by this TM is directing seat leakage from RC13A and RC13B to the Pressurizer Quench Tank via the 3/4" stainless steel tubing drain line for each valve. It should also be noted that USAR 5.2.4.7 describes temperature indication as sensed by RC12-2 and RC12-3 and increasing Quench Tank level as indicators of Pressurizer Code Safety Valve seat leakage. It is noted that this TM defeats the confirmation of indication provided by an increasing Quench Tank Level.

 Section 5.0 - Effects on Safety: in part, The function of indication of seat leakage to control room personnel is degraded by this TM (i.e. the ability to confirm an increasing discharge pipe temperature with a corresponding increase in Quench Tank level has been defeated). As stated, the control room will have indication of seat leakage via computer points T770 and T771 (sensing elements TE RC 12-2 and TE RC 12-3. Also, After the installation of the TM, pressurizer code safety valve seat leakage will become a portion of the unidentified leakage and will be quantified by the RCS inventory balance.

USAR Section 5.2.4.7.b, states that Pressurizer relief valves -reactor coolant inventory reduction as a result of seat leakage through the pressurizer pilot-operated relief valve may be identified by either of these indications: 1) high temperature downstream of the relief valve; 2) increased level in the pressurizer quench tank that provides and indication of the magnitude of the leak. Option 1 is still available with the temporary modification installed.

This temporary modification was removed on May 8, 1999.

Reference Material - Safety Evaluation 98-0045, Temporary Modification 98-0036, and the Davis-Besse Nuclear Power Station Updated Safety Analysis Report RAM Item No. - SUP-14 Closed: Y Description of Issue - Strategic Performance Area(s) Identification: Inspection Requirements 02.02, 02.07, and 02.08 should always be performed regardless of the strategic performance areas selected for review.

Description of Resolution - This item is redundant to the individual SUP items contained in SUPs 15 through 20, 104 and 105. As such, this item is being closed administratively as a duplicate.

Reference Material - None.

RAM Item No. - SUP-21 Closed: Y

Description of Issue - Assessment of Performance in the Reactor Safety Strategic Performance Area: Inspection Preparation: Develop an information base to allow the review of the effectiveness of corrective actions by compiling performance information from the licensees corrective action program, audits, self-assessments, licensee event reports (LERs), and the inspection report record (both the inspection reports and the PIM) for the time period.

Description of Resolution - The database being used is the inspection plan for the Corrective Action Team Inspection (CATI), which includes several examples of condition reports, corrective action documents, unresolved items, and LERs. The CATI inspection plan is attached to the March 14, 2003, meeting minutes of the Davis-Besse Oversight Panel.

Reference Material - Minutes for the March 14, 2003, Davis-Besse Oversight Panel Meeting.

RAM Item No. - SUP-22 Closed: Y Description of Issue - Assessment of Performance in the Reactor Safety Strategic Performance Area: Inspection Preparation: Develop an information base to allow the review of the effectiveness of corrective actions by reviewing the compiled information from SUP-21 and sort the issues by the key attributes.

Description of Resolution - Same basis for closure as SUP-21.

Reference Material - Minutes for the March 14, 2003, Davis-Besse Oversight Panel Meeting.

RAM Item No. - SUP-24 Closed: Y Description of Issue - Assessment of Performance in the Reactor Safety Strategic Performance Area: Inspection Preparation: Perform the following inspection requirements for each key attribute focusing on the selected system.

Description of Resolution - This item is redundant to the individual SUP items contained in SUPs 25 through 57. As such, this item is being closed administratively as a duplicate.

Reference Material - None.

RAM Item No. - LER-01 Closed: Y Description of Issue - Review and Evaluate Main Steam Safety Valve Setpoints Greater than Allowable Description of Resolution - Appropriate actions per Technical Specification (TS) 3.7.1.1 were taken, until each valve was adjusted and demonstrated proper operation within the allowable band. Licensee corrective action implementation is documented as closeout to Condition Report CR 02-0502.

Reference Material - Inspection Report 02-05, which is in ADAMS as Accession No. ml022060551.

RAM Item No. - LER-03 Closed: Y Description of Issue - Review and Evaluate Fuel Movement in Spent Fuel Pool without required door LER.

Description of Resolution - This LER documents a situation where the emergency ventilation system was inoperable for approximately 37 minutes during a time when it was required to be operable. The full details of the event and associated closure are addressed in NRC Inspection Report No. 2002-005. Technical Specification 3.9.12 permits an emergency ventilation train servicing the spent fuel pool area to be considered operable when the containment equipment hatch is open and both doors of the containment personnel are open, provided at least one personnel air lock door is capable of being closed and a designated individual is available immediately outside the personnel airlock to close the door. During that 37 minutes, no designated individual was immediately available to close the personnel airlock door. Although this is a violation of TS 3.9.12, it constitutes a violation of minor significance that is not subject to enforcement action in accordance with Section IV of the Enforcement Policy. Licensee corrective action implementation is documented as closeout to Condition Report CR 02-1199.

Reference Material - Inspection Report 02-05, which is in ADAMS as Accession No. ml022060551.

RAM Item No. - LER-04 Closed: Y Description of Issue - Review and Evaluate containment isolation closure requirements for RCP Seal Injection Valves LER. As a result of this condition, during postulated accident conditions, a potential for uncontrolled radioactive leakage outside containment could be created. This condition has apparently existed since original plant construction, and is a violation of Technical Specification 3.6.3.1 for Modes 1-4. In addition, the valves were determined to be installed inconsistent with design assumptions. The causes of these conditions are less than adequate design interface communication and design control.

Description of Resolution - The licensee identified that the pressure regulating valve setpoint for the reactor coolant pump (RCP) seal injection was inadequate to ensure closure of the valves upon receipt of a containment isolation signal. The condition apparently existed since original plant construction. Downstream of these isolation valves are check valves that are designed to prevent flow out of the reactor coolant system, thereby isolating the flow path regardless of whether the RCP seal injection valves are closed. The test history of the check valves was determined to be highly reliable and had no test failures in the past 10 years. The regional SRA performed a Phase 3 assessment and determined that the issue had very low safety significance (green) due to the low initiating event frequency of an interfacing system loss of coolant accident (ISLOCA), 1E-7, coupled with the check valves failure probability to prevent a potential ISLOCA if the RCP seal injection valve failed. The SRA also reviewed the licensees risk assessment and determined that the calculation was conservative given the

assumptions used. The licensees analysis determined that the change in core damage frequency was in the 1E-8 range.

This LER was originally discussed in IR 50-346/02-10 and considered to be an Unresolved Item (URI) (URI 50-346/02-10-2), pending a formal evaluation of the risk imposed by this design issue.

LER 50-346/2002-004-00; Containment Isolation Closure Requirements for RCP Seal Injection Valves MU66A-D, was closed and URI 50-346/02-10-2 were closed in inspection report 2002-017. A licensee-identified violation associated with this issue is discussed in Section 4OA7 of the inspection report.

Licensee corrective action implementation is documented as closeout to Condition Report CR 02-02254, RCP Seal Injection Air Operated Valves Will Not Perform Safety Function.

Reference Material - NRC Inspection Report No. 2002-017, which is in ADAMS as accession No. ml023430380.

RAM Item No. - URI-11 Closed: Y Description of Issue - Containment Isolation Closure Requirements for RCP Seal Injection Valves MU66AD. As a result of this condition, during postulated accident conditions, a potential for uncontrolled radioactive leakage outside containment could be created. This condition has apparently existed since original plant construction, and is a violation of Technical Specification 3.6.3.1 for Modes 1-4. In addition, the valves were determined to be installed inconsistent with design assumptions. The causes of these conditions are less than adequate design interface communication and design control.

Description of Resolution - This issue is the same as for LER-04. See description of resolution for LER-04 for closure of this URI.

Reference Material - NRC Inspection Report No. 2002-017, which is in ADAMS as accession No. ml023430380.

RAM Item No. - URI-41 Closed: Y Description of Issue - Inappropriate Licensee Notification of NRC Inspector Activity and failure of licensee personnel to enforce an obvious OSHA safety deficiency.

Description of Resolution - The inspector informed licensee management that employees who are aware of safety requirements should enforce those requirements when deficiencies are observed. Additionally, licensee employees should not warn other licensee employees of the NRC inspectors presence as it could leave impression that behavior of licensee individuals was dependent on whether or not the NRC inspector was watching a given activity. Although 10 CFR 50.70(b)(4) requires, in part, that the arrival and presence of the NRC inspector is not announced or otherwise communicated by its employees or contractors to other persons at the

facility unless specifically requested by the NRC inspector, the inspectors determined that the advance notice in this case was not a violation of regulatory requirements. "[10 CFR] Part 50 Statement of Considerations," October 25, 1988, states that "The intent of this rule is to prevent site and contractor personnel from widespread dissemination . . . of the presence of an NRC inspector. It further states that " . . . the NRC expects to reserve enforcement action for significant intentional violations of the rule." The inspectors determined that there was no widespread dissemination of the presence of the NRC inspectors.

A licensee mechanical maintenance person observed the NRC inspectors signing in at the auxiliary building radiation protection access point prior to entering containment. The same person warned the two other licensee employees of the NRC inspectors in containment. In addition, the inspectors determined that there was no significant intentional violation of the rule.

The licensee reviewed General Employee Training and found no specific reference to 10 CFR 50.70(b)(4). Because the mechanical maintenance person was not trained on the regulation the inspectors determined that there was not a significant intentional violation of the rule. The licensee took action to include the regulation in General Employee Training.

This issue was discussed in inspection report 2002-017 and documented in the licensee corrective action program as CR 02-9278. In response to this issue, the licensee conducted an independent investigation of the event and conducted site wide training on the requirements of 10CFR50.70(b)(4). The training was completed in November 2002 and the inspectors were briefed on the results of the licensee investigation.

Reference Material - See Inspection Report No. 2002-017, which is in ADAMS at accession no. ml023430380.

RAM Item No. - NCV-01 Closed: Y Description of Issue - Failure to have procedural guidance to control the construction of scaffolding in a manner that would assure proper operation of ventilation of safety equipment.

During a run of EDG 2, scaffolding restricted air circulation and produced a high temperature condition on the EDG.

Description of Resolution - This issue was identified and resolved in NRC Inspection Report No. 2002-010. Due to plant conditions at the time, no fuel in the reactor pressure vessel and no fuel movement in progress, the inspectors concluded that the finding was a Non-SDP green finding consistent with guidance in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports." This issue was placed in the licensee corrective action program as Condition Report CR 02-03570. The completed corrective action associated with this issue was a change request to procedure DB-MS-01637, Scaffolding Erection and Removal, to ensure consideration of the impact of scaffolding erection on ventilation system heat removal capability.

Reference Material - Licensee Condition Report No. 02-03570 and NRC Inspection Report No. 2002-010, which is in ADAMS at accession no. ml023030585.