ML041270225

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Minutes of Internal Meeting of the Davis-Besse Oversight Panel
ML041270225
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 04/29/2004
From: Grobe J
NRC/RGN-III
To:
NRC/RGN-III
References
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April 29, 2004 MEMORANDUM TO: Davis-Besse Oversight Panel FROM: John A. Grobe, Chairman, Davis-Besse Oversight Panel /RA/

SUBJECT:

MINUTES OF INTERNAL MEETING OF THE DAVIS-BESSE OVERSIGHT PANEL 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 February 12, 2004. Attached for your information are the minutes from the internal meeting of the Davis-Besse Oversight Panel, and the approved RAM Closure Forms.

Attachments: As stated cc w/att: D. Weaver, OEDO J. Caldwell, RIII G. Grant, RIII S. Reynolds, DRP B. Clayton, EICS DB0350

DOCUMENT NAME: C:\ORPCheckout\FileNET\ML041270225.wpd To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE RIII RIII RIII NAME RBaker/rdb CLipa JGrobe DATE 04/22/04 04/28/04 04/29/04 OFFICIAL RECORD COPY

MEETING MINUTES: Internal IMC 0350 Oversight Panel Meeting Davis-Besse Nuclear Power Station DATE: February 12, 2004 TIME: 6:00 a.m. Central ATTENDEES:

J. Grobe W. Ruland C. Lipa A. Mendiola R. Baker J. Hopkins S. Thomas Agenda Items:

1. Discuss/Approve Todays Agenda The Panel approved the agenda, which was a focused topic agenda to discuss RAM items ready for closure. R. Baker led a review of the RAM items presented for closure, and THE RESTART ACTION MATRIX ITEMS THAT THE PANEL APPROVED FOR CLOSURE ARE ATTACHED TO THESE MINUTES.

RAM Items Approved for Closure at Panel February 12, 2004 RAM Item No. - C-29 Closed: Y Description of Issue - Observe and evaluate the control rod drive testing to ensure no leakage and adequate scram times, using the new NRC Inspection Procedure 71007, Reactor Vessel Head Replacement Inspection, for guidance.

Description of Resolution - Inspection Report 05000346/2003023 documented inspection during reactor coolant system leak testing activities. The inspection included walkdowns of the reactor coolant system while at normal operating pressure as well as detailed evaluation of your inspections of the reactor vessel bottom head and closure head penetrations, and control rod drive mechanism flange connections following the 7 day pressure holding period. As a result of these pressure test activities, we have reasonable assurance that there are no pressure boundary leaks in the reactor coolant system.

Inspection Report 05000346/2004002 documents inspection of DB-SC-03270, Control Rod Assembly Insertion Time Test. This activity was observed to evaluate proper control rod movement and reactor vessel head alignment. This test was successfully completed on February 10, 2004. The inspection and testing efforts performed for closure of Restart Checklist Item 2.a, Reactor Pressure Vessel Head Replacement, exceeds guidance provided in NRC Inspection Procedure 71007.

Reference Material - NRC Inspection Report Nos. 05000346/2003023 (ADAMS Accession No.

ml033421074); 05000346/2004002.

RAM Item No. - L-23 Closed: Y Date of Letter - 07/03/02 Author - UCS Description of Issue - Why did the NRC "accept" the boric acid corrosion program at D-B after determining that over 20% of the program was "unsatisfactory"?

Description of Resolution - The NRC staff and a consultant audited Davis-Besse in September 1989, with respect to the licensee's response to GL 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." By letter dated February 8, 1990, the NRC issued its evaluation, which contained the following statement; "The purpose of this letter is to advise you that our audit of your boric acid corrosion prevention program has resulted in an acceptable finding and we now consider this issue to be closed."

The consultant's report, NUREG/CR-5576, "Survey of Boric Acid Corrosion of Carbon Steel Components in Nuclear Plants," was published in June 1990. In that report, Davis-Besse, although unnamed, received a rating of 2 (unsatisfactory, with recommended improvement actions) in 2 of the 9 categories rated. The Executive Summary of NUREG/CR-5576 contains the following statement, "All of the audited licensees satisfactorily addressed the four specified requirements of Generic Letter 88-05."

In NUREG/CR-5576, the lowest rating of the rating scale was 1 (noncompliance). So although, Davis-Besse received two ratings of 2 (unsatisfactory, with recommended improvement actions), there were no ratings of 1 (noncompliance). Overall the report author concluded that the licensees program was adequate, and that receiving a 2 rating in one of the four GL 88-05 items did not lead the report author to automatically conclude that a licensee was unsatisfactory with regard to addressing the item.

Based on the above, the NRC's letter of February 8, 1990, and the consultant's report were consistent in their conclusions that the licensee had satisfactorily addressed GL 88-05.

Reference Material - NUREG/CR-5576, "Survey of Boric Acid Corrosion of Carbon Steel Components in Nuclear Plants,"; NRC's letter of February 8, 1990.

RAM Item No. - LER-09 Closed: Y Description of Issue - On November 29, 2002, with the reactor defueled, it was discovered that the thermal sleeve connected to the 2-2 HPI/makeup nozzle had an axial crack. Inspection of the 2-1 HPI/makeup thermal sleeve also revealed a cracked thermal sleeve. No cracking was observed during the inspection of the remaining two HPI thermal sleeves.

Restart Checklist Item - 2.e Description of Resolution - The remedial action was to replace the thermal sleeves. Inservice inspection procedures were developed to ensure enhanced inspection techniques would be used in the future to verify the integrity of the HPI/makeup thermal sleeves. The licensee stated that the visual inspections will include the use of high resolution video equipment and verification that the video equipment was applied in accordance with ASME Section XI, sub-article IWA 2210, Visual Exam for VT-1 Examination. The licensee stated that the frequency of inspection would be every other refueling outage. This LER is closed in NRC Inspection Report No. 50-346/03-10.

Reference Material - NRC Inspection Report 50-346/03-10 (ADAMS Accession No.

ml040680070).

RAM Item No. - LER-14 Closed: Y Description of Issue - LER 2003-005, Revision 00 and 01, Containment Gas Analyzer Inoperability Due To Isolation of Cooling Water Description of Resolution - The licensees initial submittal of this LER discussed a condition where the component cooling water (CCW) isolation valves on the inlet and outlet to the heat exchangers located in each of the two Containment Gas Analyzers Systems (CGAS) were found stuck shut. This condition rendered the CGAS incapable of performing its design function. This issue was evaluated by the inspectors and documented in Inspection Report No. 50-346/03-17.

Supplement 01 to this LER, dated January 23, 2004, described the following two additional issues that were identified by the licensee during the extent of condition evaluation, that directly impacted the proper operation of the hydrogen gas analyzers:

  • The instrument air supplied to the moisture trap drain check valve associated with the hydrogen analyzers heat exchanger was non-safety grade. Post accident, this air supply would not have been available and the drain valve would not have functioned.

Additionally, the regulator which supplied the air to the drain valve was set at too low for the drain valve to operate as designed.

  • The moisture traps potentially contaminated condensate, via the drain valve, would flow to a floor drain in a room not served by the emergency ventilation system. This constituted a potential containment bypass pathway.

The licensee addressed the first issue by eliminating the reliance of the drain check valve on instrument air by replacing the air operated drain valves with solenoid operated valves, powered by independent essential power sources. The second issue was corrected by routing the potentially contaminated condensate to an existing ECCS floor drain.

The LER is closed in Inspection Report No. 50-346/04-02.

Reference Material - NRC Inspection Report Nos. 50-346/03-17 (ADAMS Accession No.

ml032721592), and 50-346/04-02.

RAM Item No. - NCV-6 Closed: Y Description of Issue - In 2002, SSDI team identified NCV-6 for failing to correctly translate the design basis requirements for sizing of the safety-related backup air supplies for containment isolation valves SW-1356, SW-1357, and SW-1358 into the design. The licensees corrective action was to install new accumulators sized to hold the valves closed.

Description of Resolution - The team reviewed several revisions of calculation C-ME-011.06-007 which sized the new accumulators. The team identified errors in the calculation which required the calculation to be revised. For example, in Revisions 0 and 1 of the calculation, the new accumulators were intended to be filled with air as the licensee thought the valves only had to remain closed for 30 minutes. The licensee initially did not recognize that the valves had a containment isolation design function which required the valves to remain closed for 30 days.

Following the team's questions, the licensee changed the design to require that the new accumulators be filled with nitrogen rather than air. In the last revision reviewed, the calculation erroneously used the ideal gas law equations when sizing the nitrogen bottles without consideration of the compressibility of nitrogen at a pressure of 2000 psig. Additional examples of lack of rigor and inadequate assumptions were identified. The licensee revised the calculation to correct the errors identified by the team. This item is resolved.

A NCV of 10 CFR Part 50, Appendix B, Criterion III, having very low safety significance (Green) was issued in NRC Inspection Report 05000346/2003010.

Reference Material - NRC Inspection Report 05000346/2003010, Sections 4OA3(3)b.4 (ADAMS Accession No. ml040680070) and NCV 05000346/2002014-01a.

RAM Item No. - NCV-7 Closed: Y Description of Issue - NCV-7 was issued by the NRC during the 2002 SSDI to document that there were no provisions to blow down the SW containment isolation valve accumulators although USAR Section 9.3.1.5 stated that the accumulators contained a provision to allow removal of excessive moisture.

Description of Resolution - The team determined that, due to the change in accumulator medium from air to nitrogen, that there was no longer any need for blowdown provisions.

Reference Material - NRC Inspection Report 05000346/2003010, Sections 4OA3(3)b.5 (ADAMS Accession No. ml040680070) and NCV 05000346/2002014-01b.

RAM Item No. - NCV-08 Closed: Y Description of Issue - The licensee initiated CR 02-07766 to address the issue that the trip set point specified in calculation C-EE-004.01-049 was greater than the TS allowable value.

Therefore, the postulated TS allowable value could be violated for plant operating conditions where the voltage was just above the relay set point value.

Description of Resolution - The team reviewed the issue and determined that the new calculation, C-EE-015.03-008, which utilized the ETAP program, properly addressed all issues included in the CR. Therefore, the corrective actions to this issue were deemed acceptable, and this item is resolved.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-01i.

RAM Item No. - NCV-11 Closed: Y Description of Issue - NCV-11 addressed the fact that a surveillance test did not demonstrate that worst-case post-accident conditions were bounded for the CAC discharge valves in the SW system.

Description of Resolution - The licensee was replacing these valves, due to a number of problems with them. The proposed corrective actions appeared to include appropriate acceptance criteria. The team identified a concern with the original evaluation and corrective action wording in CR 02-07781. The licensees procedure did not declare the valves inoperable and write a CR if the valves failed the valve closure test. This issue was not originally addressed in the licensees corrective actions. However, when it was brought to the licensees

attention, appropriate changes were made in the procedure to address declaring the valve inoperable and writing CRs when necessary. This item is adequately resolved.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-02a.

RAM Item No. - NCV-13 Closed: Y Description of Issue - The licensee failed to perform a surveillance in accordance with TS 4.5.2.H for HPI pump following maintenance.

Description of Resolution - The licensee requested a TS amendment (No. 256) to relocate the surveillance requirement pertaining to flow balance testing of the HPI and LPI subsystems following system modifications to the technical requirement manual. The amendment added ECCS pump operability conditions to the TS. The new surveillance requirement would require verifying each ECCS pumps developed head to be greater than or equal to the required developed head, when tested pursuant to TS 4.0.5 with regards to inservice testing requirements of the ASME Code. The team had no further concerns, did not identify other new issues, and has evaluated this item is resolved.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-04.

RAM Item No. - SUP-15 Closed: Y Description of Issue - Determine whether licensee evaluations of, and corrective actions to, significant performance deficiencies have been sufficient to correct the deficiencies and prevent recurrence.

Description of Resolution - The team determined that the licensee's program for identifying, prioritizing, evaluating, and correcting performance deficiencies was adequate. However, the licensees actions were repeatedly insufficient to identify the issue and prevent recurrence. The licensees evaluations were inadequate and were based upon preconceived conclusions. The corrective actions identified from the inadequate evaluations were also inadequate. Also, few of the corrective actions had been in place long enough for either the team or the licensee to assess the overall effectiveness of the implemented corrective actions.

The CATI identified violations of 10 CFR 50, Appendix B, Criterion III and XVI, which involved the licensee not taking corrective actions to resolve previously documented non-cited violations.

Additionally, the CATI identified numerous violations of very low safety significance (Green) and a Severity Level IV violation (relating to 10 CFR 50.59).

The team identified some improvements which the licensee had made in the CAP. Examples included the revised CAP procedure and the newly established CR analyst positions.

The licensee recognizing the extent of the inspection findings, developed improvement plans to address the identified deficiencies and provide additional barriers to ensure that engineering products were of acceptable quality. These plans were described in the licensees Operational Improvement Plan, Operating Cycle 14. The plan includes the areas of concern identified by the team.

Reference Material -NRC Inspection Report 05000346/2003010, Section 4OA2(1).b (ADAMS Accession No. ml040680070).

RAM Item No. - SUP-36 Closed: Y Description of Issue - IP 95003; 02.03.c.2.b: Review specific problem areas and issues identified by inspections to determine if concerns exist in training and qualifications.

Description of Resolution - Inspection activities, primarily in the area of Operations were documented in several inspection reports (05000346/2003002, 05000346/2003011, 05000346/2003017, 05000346/2003018, 05000346/2003022, 05000346/2003025). These inspection activities included assessing the biennial written examination and annual operating test results, observations of just-in-time training conducted prior to important plant evolutions, observations of operator performance during annual requalification simulator examinations, and the direct application of the training as demonstrated performance by operator in the plant.

The inspectors also reviewed training materials, developed to address demonstrated maintenance performance deficiencies, which were presented to maintenance personnel. The inspectors believe that the efforts in this area by some specific mid-level maintenance supervisors, were a key factor in the improvement of the maintenance department in the areas of work quality and procedure adherence.

The inspectors determined that, with some minor discrepancies, the licensee conducted operator training at an acceptable level to provide operators with the knowledge necessary to properly operate the plant systems. This item is resolved.

Reference Material - NRC Inspection Report Nos.: 50-346/03-02 (ADAMS Accession No.

ml030690302); 50-346/03-11 (ADAMS Accession No. ml040360097); 50-346/03-17 (ADAMS Accession No. ml032721592); 50-346/03-18 (ADAMS Accession No. ml033080433);

50-346/03-22 (ADAMS Accession No. ml033570081); 50-346/03-25 (ADAMS Accession No.

ml040290768).

RAM Item No. - SUP-42 Closed: Y Description of Issue - IP 95003; 02.03.f.3: Evaluate the quality of procedures and as applicable, determine the adequacy of the procedure development and revision process.

Description of Resolution - During the past year, the licensee has had some issues with procedure quality and procedure adherence and/or implementation. These issue were documented in several inspection reports.

Specific examples of poor procedure quality included:

  • A self-revealing Non-Cited Violation of very low safety significance was identified for inadequate component restoration instructions contained in DB-SC-03122, SFAS Component Testing Procedure, Revision 01. This resulted in the inadvertent operation, on separate occasions, of Borated Water Storage Tank Outlet Valves DH7A and DH7B during Safety Feature Actuation System (SFAS) individual component testing restoration activities for Core Flooding Tank to Sampling System Valve CF1545 and Nitrogen System to Containment Isolation Valve NN236. (Inspection Report 2003-013)
  • A self-revealing Non-Cited Violation of very low safety significance was identified for failing to provide adequate procedural guidance for tightening fasteners internal to the high pressure injection pump. As a direct result, five socket head cap screws, located near the discharge of the pump, failed during pump testing. (Inspection Report 2003-015)
  • A self-revealing Non-Cited Violation of very low safety significance was identified when it was determined that the procedure for testing the response time of the auxiliary feedwater pump 1 turbine did not adequately describe the acceptance criteria for successful completion of the test. (Inspection Report 2003-018)
  • An NRC identified Non-Cited Violation of very low safety significance was identified when the inspectors discovered that procedural guidance which governed the performance of the Immediate Action Maintenance (IAM) process did not exist.

(Inspection Report 2003-018)

  • On September 5, 2003, during a plant heatup to establish test conditions for the reactor coolant system normal operating test, CF1B opened unexpectedly when reactor coolant system pressure increased to the valves automatic actuation set-point. (Inspection Report 2003-018; Minor Violation)

Specific examples of poor procedure adherence and/or implementation:

  • An NRC identified Non-Cited Violation of very low safety significance was identified for the failure to properly implement procedures required for performing equivalency evaluations for components being replaced in safety related equipment. This resulted in the installation of relays into the Safety Features Actuation System (SFAS) cabinets that were not electrically rated for their specific application. (Inspection Report 2003-013)
  • A self-revealing Non-Cited Violation of very low safety significance was identified for the failure to properly implement work instructions during the reinstallation of electrical conduit and the electrical termination of operating power and indication power to Loop 1 Reactor Coolant System High Point Vent Valves RC4608A and RC4608B. This resulted in the electrical power for each valve being swapped. Inspection Report 2003-013)
  • A self-revealing Non-Cited Violation of very low safety significance was identified for failing to properly implement system procedures during the filling of the circulating water system. Since three drain valves were improperly left open during the fill, approximately three inches of water flooded the 565' elevation of the turbine building. (Inspection Report 2003-015)
  • A self-revealing Non-Cited Violation of very low safety significance was identified for failing to perform work in accordance with approved maintenance procedures during the installation of reactor coolant pump mechanical seal RTDs. As a direct result, the RTD tubing nuts were not installed to a sufficient tightness to provide a leak tight joint at normal operating pressure. (Inspection Report 2003-015)
  • An NRC identified Non-Cited Violation of very low safety significance was identified when the inspectors discovered a significant amount of loose material in the

containment building, subsequent to a final closeout inspection performed by senior licensee management. (Inspection Report 2003-018)

  • An NRC identified Non-Cited Violation of very low safety significance was identified when the inspectors discovered that Operations management inappropriately authorized the performance of the IAM process to perform adjustments on 1 turbine driven auxiliary feedwater pump governor. (Inspection Report 2003-018)
  • While performing Section 4.2 of DB-PF-03080, AFW Check Valves AF1, AF2, AF15, and AF 16 Reverse Flow Tests, Revision 00, the initial system conditions, using the guidance stated in the procedure, could not be established to perform the test. To correct this condition, the test leader attempted to vent the upstream pressure seen by the valves. Steps for this venting were not in the procedure and the specific approval was not obtained from control room staff prior to manipulating the vent valves.

(Inspection Report 2003-018; Minor Violation)

  • While attempting to establish additional turbine plant cooling water flow through the generator hydrogen coolers utilizing procedure DB-OP-06263, Turbine Plant Cooling Water System, Revision 03, a spill of approximately 80 gallons occurred due to vent and drains valves associated with the generator hydrogen coolers being inappropriately left open. (Inspection Report 2003-018; Minor Violation)

The resident staff has reviewed the corrective actions for each of these issues and found them to be adequate. The inspectors have noted improving trends in the areas of procedure quality and implementation, as evidenced by:

  • the increased willingness of maintenance and operations personnel to submit conditions reports for deficient maintenance procedures;
  • a decrease in the number of maintenance procedural non-compliances during work activities;
  • significant improvements to integrated operational procedures (i.e., heatup, startup, cooldown); and
  • a significant decrease (since the end of September, 2003) in the number of procedure related errors.

Based on the review of the corrective actions associated with the performance issues discussed above and the current licensee trends in the are of procedure quality, the inspectors determined that the licensees performance in this area to be adequate.

Reference Material - NRC Inspection Report Nos.: 50-346/03-13 (ADAMS Accession No.

ml031680985); 50-346/03-15 (ADAMS Accession No. ml032120360); 50-346/03-17 (ADAMS Accession No. ml032721592); 50-346/03-18 (ADAMS Accession No. ml033080433).

RAM Item No. - SUP-44 Closed: Y Description of Issue - Assessment of Performance in the Reactor Safety Strategic Performance Area: Key Attribute - Equipment Performance: Assess the effectiveness of corrective actions for deficiencies involving equipment performance, including equipment designated for increased monitoring via implementation of the Maintenance Rule.

Description of Resolution - The inspector reviewed the reference materials and interviewed the Maintenance Rule (MR) Program owner. The MR program received a comprehensive examination by the Program Review Board (PRB) as documented in reference 4. This covered the program purpose, ownership, scope, deviations from regulatory basis documents, implementation, performance indicators, recent improvements, assessments, and outstanding items assigned by the PRB. The inspector noted that the chairman of the PRB was especially qualified to evaluate the MR program since he was the senior operations engineer in the NRR MR section from 1995 until 2001. The PRB concluded that the program was in a condition ready to support restart and operation but identified a number of areas where the program needed improvement. Non-restart condition reports were written to document this areas. They included:

  • upgrade scoping sheet descriptions for MR functions
  • compare cycle 11 and 12 functional failures to cycle 13 functional failures for trending purposes
  • upgrade the risk matrix tool
  • develop and implement process to incorporate risk level changes from procedure changes into the Safety Monitor Program
  • change to focus of the MR program from raw regulatory compliance to using the program to bring attention to deteriorating physical plant conditions and improving overall equipment reliability and unavailability
  • the MR program did not have support from the total plant organization and needed to gain better support from Operations, Maintenance, PRA/PSA, Engineering
  • MR program needed to establish metrics so as to understand how performance and condition monitoring are being evaluated, and with what frequency, so as to ensure compliance with evaluation under (a)(3) of the MR Rule.

The inspector interviewed the MR Program owner to assess his knowledge and ownership of the program, to learn his assessment of the program, and his plans for improvements. While he had only been the MR owner for approximately a year, he was knowledgeable and appeared committed to the program. He stated that plant engineers looked at the program favorably since an (a)(1) classification served to draw attention and resources to the system. This is consistent with the views of system engineers interviewed by the inspector during the MR Baseline Inspection in January 1997.

Based on the above, the inspector considered the MR program acceptable for plant operations.

Reference Material -10 CFR 50.65 - Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants; DB-PF-0003, Maintenance Rule; Maintenance Rule Program Manual; Program Readiness Baseline Assessment Package for Maintenance Rule Program, Revision 00, February 4, 2003 RAM Item No. - SUP-50 Closed: Y Description of Issue - IP 95003; Section 02.03.f.3: Determine that the work control process uses risk appropriately during planning and scheduling of maintenance and surveillance testing activities and the control of emergent work.

Description of Resolution - During 2003, the resident inspectors evaluated 21 samples utilizing the Maintenance Risk and Emergent Work inspection procedure [71111.13]. The inspectors reviewed the licensees response to risk significant activities. Activities chosen were based on their potential impact on increasing overall plant risk. The inspections verified the planning, control, and performance of the work were done in a manner to control overall plant risk and minimize the duration where practical, and that contingency plans were in place, where appropriate. The licensees daily configuration risk assessments, observations of shift turnover meetings and observations of daily plant status meetings were evaluated by the inspectors to verify that the equipment configurations had been properly listed, that protected equipment had been identified and was being controlled where appropriate, and that significant aspects of plant risk were being communicated to the necessary personnel.

During these inspections, the inspectors identified a finding of very low safety significance when the Operations management inappropriately authorized the performance of the Immediate Action Maintenance Process to perform adjustments on turbine driven auxiliary feedwater pump 1 governor. Additionally, another finding of very low safety significance was identified when the inspectors discovered that procedural guidance which governed the performance of the Immediate Action Maintenance process did not exist. These findings were documented in Inspection Report No. 50-346/03-18. The corrective actions for these performance deficiencies were evaluated by the inspectors and found to be acceptable.

Based on these documented inspection activities and day-to-day observations of how the licensee incorporates risk insights into work scheduling, the inspectors determined that the licensee performs this function in a satisfactory manner.

Reference Material - NRC Inspection Report No. 50-346/03-18 (ADAMS Accession No.

ml033080433).

RAM Item No. - SUP-62 Closed: Y Description of Issue - IP 71007, Reactor Vessel Head Replacement Inspection, step 02.05, provides guidance for post-installation verification and testing inspections. This step recommends that selected inspections be conducted in the following areas: containment testing; licensees post-installation inspections and verifications program and its implementation; RCS leakage testing; post installation equipment testing.

Description of Resolution - The following inspection activities were sufficient in scope to close this item.

Reactor Vessel Removal and Replacement - The physical removal of the old reactor vessel head from containment and the movement of the new reactor vessel head into containment were observed as routine resident plant status activities and was not specifically documented in an inspection report.

Inspection reports 05000346/2002010 and 05000346/2003017 documented radiological inspections associated with head replacement activities. Specific inspection activities included:

  • walkdowns of selected portions of the radiologically restricted area, including areas within the Auxiliary and Containment Buildings where significant radiological work involving the reactor head and containment breach was occurring;
  • observed work occurring both inside and outside of the Containment Building including preparation for the reactor head moves and Containment Building breach;
  • walkdowns of areas outside of the Containment Building where equipment for making the Containment breach was operating to verify that controls for containing radioactive materials generated in the breach process were adequate;
  • reviewed the reactor head encapsulation process to verify that contamination control and radiological shielding were adequate to minimize dose to workers and to meet 10 CFR and 49 CFR requirements for the eventual transportation of the reactor head to a burial site; and
  • observed aspects of the preparation of a shipment of the reactor head including the shipping documentation.

Design and Planning/Reactor Vessel Head Inspection - Inspection Report 05000346/2002007 documented review of the non-destructive examinations performed on the replacement head welds that occurred at the Midland Michigan site and the American Society of Mechanical Engineers (ASME) Code data packages for the replacement head. Our inspection concluded that adequate records were assembled to ensure that the replacement head was designed and fabricated in conformance with ASME Code requirements and that the original ASME Code Section III N-stamp remained valid.

Containment Vessel Restoration - Inspection Report 05000346/2002007 documented that:

  • the engineering evaluation associated with construction of the temporary containment access opening considered appropriate loads and demonstrated that stress in the containment shell materials would not exceed design limits;
  • the temporary containment vessel opening was restored such that the original ASME Code construction requirements were maintained;
  • the work activities to construct and restore the temporary containment opening and closure occurred in a controlled manner and in accordance with procedure requirements; and
  • that the licensee managers demonstrated an active oversight role for the control of the contractors on the containment building temporary construction opening.

Inspection Report 05000346/2003005 documented that:

  • based on the results of the containment integrated leak rate check, containment integrity had been restored where the containment had been opened for replacement of the reactor head.

Based on the results of these two inspection activities, the licensees efforts to construct a temporary containment access, restoration of the temporary access following reactor head movement into containment, and subsequent leak testing were adequate.

Post Installation Testing - Inspection Report 05000346/2003023 documented inspection during reactor coolant system leak testing activities. The inspection included walkdowns of the reactor coolant system while at normal operating pressure as well as detailed evaluation of your inspections of the reactor vessel bottom head and closure head penetrations, and control rod

drive mechanism flange connections following the 7 day pressure holding period. The results of these pressure test activities provide reasonable assurance that there are no pressure boundary leaks in the reactor coolant system.

Inspection Report 05000346/2004002 documented inspection of the performance of surveillance test DB-SC-03270, Control Rod Assembly Insertion Time Test. This activity was observed to evaluate proper control rod movement and reactor vessel head alignment. This test was successfully completed on February 10, 2004.

Reference Material - NRC Inspection Report Nos. 50-346/02-07 (ADAMS Accession No.

ml023370100); 50-346/02-10 (ADAMS Accession No. ml023030585); 50-346/03-05 (ADAMS Accession No. ml032230339); 50-346/03-17 (ADAMS Accession No. ml032721592);

50-346/03-23 (ADAMS Accession No. ml033421074); 50-346/04-02.

RAM Item No. - URI-14 Closed: Y Description of Issue - During CAC motor replacement, the licensee identified splitting of the motor cable insulation as documented in CR 02-05459. The cable jacket and insulation to the three CAC motor high speed windings were found to be split at the ends which were normally covered by Raychem' heat shrink sleeves. The damage was observed after the Raychem' sleeves were removed for de-terminating the motors.

Description of Resolution - The NRC determined that the splitting was in fact a deep gash and the licensee subsequently determined the gash was inflicted by a contractor when removing the Raychem' sleeves with a knife. To address this concern, the licensee initiated work orders to replace the section of the high speed cable of the three CAC motors between the motor and the penetrations with an equivalent cable. The work procedures were revised and the workers received training on the revised procedures. This item is resolved. The NRC determined this issue constituted a violation of 10 CFR Appendix B, Criterion V, (failure to properly remove Raychem' splices during the CACs motor replacement), which has minor significance and is not subject to enforcement action.

Reference Material - NRC Inspection Report Nos. 50-346/02-14 (ADAMS Accession No.

ml030630314) and 50-346/03-10 (ADAMS Accession No. ml040680070).

RAM Item No. - URI-15 Closed: Y Description of Issue - Failure to include the environmental effects of a Decay Heat Removal (DHR) pump seal failure in its moderate energy line break analysis.

Description of Resolution - Following discovery, the licensee entered the issue into its corrective action program and performed the analysis. The NRC determined that the heat load caused by failure of the DHR pump seal (an additional 21,000 btu/hr) was subsequently included in calculation C-NSA-032.02-006 and that the issue was adequately resolved. A NCV of 10 CFR Part 50, Appendix B, Criterion III, having very low safety significance (Green) was issued.

Reference Material - NRC Inspection Report Nos. 50-346/02-14 (ADAMS Accession No.

ml030630314) and 50-346/03-10 (ADAMS Accession No. ml040680070).

RAM Item No. - URI-17 and URI 18 Closed: Y Description of Issue - URI-17 concerned non-conservatisms in the analysis which analyzed the heat loads in the SW pump room and the ability of the ventilation system to maintain the pump room temperatures within a required operating range.

URI-18 dealt with the effects of a postulated auxiliary steam line break in the SW pump room and whether the licensee correctly translated the USAR commitments regarding the SW pump room environmental limits into analyses that demonstrated these limits would not be violated for design basis conditions.

Description of Resolution - Both of the items were examined together during the CATI. The heat load calculation was revised and issued as Revision 4 in early 2003. At the same time, another CR, was issued because the initial CR failed to do an extent of condition review to verify the adequacy of the SW ventilation system for all operating conditions. The extent of condition review was reported to have included a walkdown of the SW pump room and review of the revised SW ventilation calculation.

Upon review of the revised calculation in 2003, the NRC noted that the summer maximum analyzed temperature in the pump house did not include the heat load contribution of the diesel driven fire pump, which was one of the deficiencies noted in the earlier revision to the calculation. This deficiency was not addressed in the new revision to the calculation, either by including it or by providing a rationale for excluding the heat load. The NRC noted that the licensee had previously had to take actions to open the diesel generator room doors and provide alternate ventilation during the summer months. The new calculation also concluded that the penthouse louvers had to be modified (blocked) for winter operation. The NRC noted that past operability had been assured for winter operation by regularly recording pump room ambient temperature.

The NRC determined that past licensee compensatory actions (both during the summer and winter months) had prevented the equipment from being inoperable. An NCV of 10 CFR Part 50, Appendix B, Criterion III was issued in NRC Inspection Report 05000346/2003010 for URI-17.

The licensee performed a calculation assessing the environmental effects of a postulated auxiliary steam line break in the SW pump room. The calculation concluded that there were no adverse effects on the equipment in the room. The NRC also noted that the licensee had initiated engineering change request to remove the auxiliary steam line from the SW pump room. The licensee stated that this modification was an enhancement which was not required.

There was no violation identified regarding RAM item URI-18.

Reference Material - NRC Inspection Report 05000346/2003010, Section 4OA3(3)b.7 (ADAMS Accession No. ml040680070) and URI 05000346/2002014-01e and 01f.

RAM Item No. - URI-19 Closed: Y Description of Issue - On September 24, 2002, the licensee issued CR 02-06893 to document an increase from 95EF to 124EF in Rooms 105 and 115 temperature as a result of an increase of SW temperature. The CR identified the need to reevaluate cable ampacity as a result of the higher room temperature.

Description of Resolution - The team discussed the ampacity issue with the licensee, and determined there actually was not an ampacity concern. Therefore, this item is considered closed.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-01g.

RAM Item No. - URI-20 Closed: Y Description of Issue - The licensee failed to have provisions in place to protect the service water pump room from flooding.

Description of Resolution - During the SSDI in 2002, the NRC identified that no procedures were in place to isolate equipment open for maintenance in the SW pump room that could flood the room in the event of high lake water level. Therefore, the NRC questioned whether the SW system was adequately protected against flooding effects that could result from high lake water levels, from internal flooding, and from other threats to the system that could result from failure of non-seismically qualified equipment, as described in the USAR.

In response to this concern, the licensee determined that operator actions were necessary in order to ensure that the USAR statements were met. In order to ensure that the operator actions occurred, several changes to operating procedures were required. These procedural actions were taken, and this item is resolved.

A NCV of 10 CFR Part 50, Appendix B, Criterion III, having very low safety significance (Green) was issued.

Reference Material - NRC Inspection Report 05000346/2003010, Section 4OA3(3)b.9 (ADAMS Accession No. ml040680070) and URI 05000346/2002014-01h.

RAM Item No. - URI-21 and NCV-9 Closed: Y Description of Issue - URI-21 concerned the use of insufficiently supported uncertainty values in the calculation for the 90 percent Undervoltage Relays.

NCV-9 dealt with non-Conservative Relay Setpoint Calculation for the 59 Percent Undervoltage Relay

Description of Resolution - These two items were examined together during the CATI. The licensee performed additional analysis to assess the impact of using insufficiently supported uncertainty values.

The design remained adequate and there was no violation identified. URI-21 which was closed in NRC Inspection Report 05000346/2003010.

The NRC reviewed ETAP calculation C-EE-015.03-008, Revision 2. The calculation properly addressed the postulated inconsistencies and non-conservative assumptions in the uncertainty analysis. Therefore, the corrective actions to NCV 9 were evaluated as acceptable.

Reference Material - NRC Inspection Report 05000346/2003010, Sections 4OA3(2)b.7 and 4OA5(1)b.2.11 (ADAMS Accession No. ml040680070) and URI 05000346/2002014-01j and 01k.

RAM Item No. - URI-22 and URI 23 Closed: Y Description of Issue - URI-22, 05000346/2002014-01l, Inadequate Calculations for Control Room Operator Dose (GDC-19) and Offsite Dose (10 CFR Part 100) Related to High Pressure Injection Pump Minimum Flow Values, regarded concerns with the dose calculations for operators and the general public following a design basis accident. The licensee failed to translate the radiological consequences of leakage from engineered safety feature components outside containment into calculations of record for post-accident control room dose and offsite boundary dose.

URI-23, 05000346/2002014-01m, Oother GDC-19 and 10 CFR Part 100 Issues, is associated with correctly translating USAR commitments regarding calculations for GDC-19 and 10 CFR Part 100 requirements. During SSDI, the NRC determined that the USAR calculated offsite dose was based on an ECCS leakage rate of 1.6 gallons per hour (gph) while the allowable leakage rate was based on 40 gph.

Description of Resolution - Both items were examined together due to their similarity during CATI. The licensee performed a preliminary calculation in the cause analysis for CR 02-07701 to determine the increase in dose in the control room from the 500 gallons deposited in the Borated Water Storage Tank (BWST). The licensee then calculated that the total offsite dose was 236.22 rem. The total control room dose was similarly for a total of 20.366 rem.

As a result of these calculations, the licensee specified post-restart corrective actions to update the Bechtel calculation of record and the USAR to incorporate these doses. Because the corrective actions had not yet been completed, the licensee had not completed a screening or evaluation under 10 CFR 50.59. The team performed a limited evaluation of the acceptability of the increased dose under 10 CFR 50.59(c)(2)(iii), "Result in more than a minimal increase in the consequences of an accident previously evaluated in the final safety analysis report (as updated)." The team reviewed the guidance provided in Nuclear Energy Institute (NEI) standard 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, which NRC endorsed in Regulatory Guide 1.187. The team concluded that the licensee had an acceptable rationale for delaying issuance of the formal calculations until after restart.

The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance. Specifically, the licensee failed to translate the postulated radiological consequences of leakage from engineered safety feature components outside containment into calculations of record for post-accident control room dose and offsite boundary dose.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070), URI 05000346/2002014-01l, and URI 05000346/2002014-01m.

RAM Item No. - URI-30 Closed: Y Description of Issue - This URI, 05000346/2002014-01t, Service Water Source Temperature Analysis for Auxiliary Feedwater, regarded the licensee failing to analyze the service water source with respect to its potentially higher temperature condition for various design basis events and the possible impact on the ability of the Auxiliary Feed Water (AFW) system to perform its safety function. Such effects could include reduced heat absorption capability for AFW injected into the SGs and inadequate cooling of AFW lubricating oil.

Description of Resolution - The licensees evaluation concluded that temperature of AFW (seismic event with long term AFW supplied by SW) was lower than the design AFW temperature of 120EF as noted in the system description. In addition, the licensee determined that AFW equipment temperature limits were greater than 120EF. Therefore, the licensee concluded that there was no discrepant condition. The team agreed with this assessment and did not identify any violation. This item is resolved.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-01t.

RAM Item No. - URI-31 Closed: Y Description of Issue - The licensee failed to consider the worst-case grid voltages for the short circuit analyses performed in support of breaker coordination.

Description of Resolution - The licensee entered the issue into their corrective action program and performed new calculations to address the issue. The team reviewed this item and determined that calculation C-EE-015.03-003 was superseded with calculation C-EE-015.03-008. The new calculation did take into account the worst-case grid voltage conditions, and no other problems were identified. A violation of 10 CFR Appendix B, Criterion III, which has minor significance was identified.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-01u.

RAM Item No. - URI-33 Closed: Y Description of Issue - The licensee failed to identify a condition where the allowable degradation of the SW pumps did not match the design basis required flow rate for the SW pumps. In particular, the pump curve was allowed to degrade by 7 percent in accordance with IST acceptance criteria, without evaluating the required design basis flow requirement.

Description of Resolution - Vendor calculations02-123 and 02-113 were performed to address all SW hydraulic issues. The allowable SW pump degradation was included in the new calculations. The team did not identify any violation and this item is closed.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-03a.

RAM Item No. - URI-35 Closed: Y Description of Issue - This was a potential nonconservative temperature measurement performed by the licensee for ultimate heat sink temperatures.

Description of Resolution - The team determined that the licensees procedures had been revised to incorporate the temperature instruments uncertainty calculation results, and that the procedures required the plant staff to take appropriate actions should it appear that the ultimate heat sink temperature was being approached (such as measuring the temperature locally with sensitive measuring and test equipment.) Therefore, the team determined that no violation existed and this issue is closed.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-03c.

RAM Item No. - URI-36 Closed: Y Description of Issue - Licensee failed in overestimating the nozzle flexibility by a factor of one thousand when analyzing the structural integrity of the connection in the SW system to the CACs.

Description of Resolution - Stress analyses concluded that the CACs were operable in the past regarding structural concerns identified in CR 02-05563. The structural report concluded that, "...Based on the lack of significance or the continued structural acceptability identified with the numerous finding associated with the CAC coil modules and their support structure, the CAC operability assessment is considered to be unaffected by the composite findings of all currently identified, structural-related CAC concerns". The team determined that the licensee appropriately used ASME Code F stress criteria in the structural analysis. This item is closed.

Reference Material - NRC Inspection Report 05000346/2003010 (ADAMS Accession No.

ml040680070) and URI 05000346/2002014-03e.

RAM Item No. - URI-38 Closed: Y Description of Issue - In November of 2002, the NRC Identified a Potential Concern for Inadequate Over-pressure Protection (OP) for the Containment Air Coolers (CACs), Decay Heat Removal (DHR) Coolers, Emergency Diesel Generator Jacket Water (EDGJW) Heat Exchangers and Associated System Piping (URI No. 06 in Inspection Report 05000346/2002014).

Description of Resolution - On January 23, 2004, the inspectors completed the on-site inspection of URI 05000346/2002014-06. This review was focused on the location of the system relief valves to ensure OP was provided for the CACs, EDGJW heat exchangers and DHR coolers under operating/design basis conditions. This review was prompted by previous NRC questions/concerns for implementation of the Code OP requirements primarily focused on the CACs. For example, the NRC had questioned the use of locked open valves between the relief valve and the Code components requiring relief protection with respect to meeting the Code requirements for positive controls and interlocks on stop valves. These specific requirements and system configurations associated with OP protection were discussed with NRC staff in the Office of NRR and no concerns for Code compliance were identified.

Specifically, the inspectors confirmed that:

C The EDGJW coolers and CACs were not Code stamped vessels and thus did not have component level design requirements governing OP protection. The OP protection for the CACs was provided by pressure relief devices for the service water system in which the CACs were installed.

C The DHR coolers were Code stamped vessels, which had component level OP protection requirements from the original design Code (ASME Code,Section III and VIII, 1968 Edition). The inspectors confirmed that the configuration and location of the system OP protection devices was consistent with these requirements.

C For the component cooling water, service water and decay heat removal piping systems which contained these components, the applicable design Code was the ASME Code,Section III, 1971 Edition. This design Code contained specific requirements associated with the location, capacity and types of relief protection required. The inspectors confirmed that the configuration and location of the system OP protection devices was consistent with these requirements for the piping sections containing these components.

For these systems and components, the licensee had not produced a written document that explicitly identified how the applicable OP protection requirements from the design Codes were implemented. Without a written record describing how the Code OP protection requirements were implemented, the inspectors were concerned that changes to the plant design or operation could place these systems/components outside the Code design basis. For example a change in plant operating lineups or system components could render the Code OP protection strategy ineffective and ultimately result in damaged equipment. Based upon this observation, the licensee implemented actions (CR 04-0052) to document the OP protection strategy for these systems and components in controlled safety-related calculations.

In conclusion, the inspectors did not identify any system normal or emergency operating configurations or lineups that would result in isolating the CACs, EDGJW coolers and DHR coolers from OP protection devices, without considering these components and associated

piping systems inoperable. Further, no deviations from applicable Code requirements were identified with respect to location of relief protection devices for these components. Therefore, URI 50-346/2002-014-06 is considered closed.

Reference Material - Inspection Report No. 50-346/04-02.

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