IR 05000247/2004003

IR 05000247-04-003, on 11/17-21 and 12/08 - /2003, Indian Point Energy Center, Unit 2; Biennial Baseline Inspection of Problem Identification and Resolution; Problem Identification and Resolution

ML040360248
Person / Time
Site:	Indian Point
Issue date:	02/04/2004
From:	Lanning W Division of Reactor Safety I
To:	Dacimo F Entergy Nuclear Northeast
Shared Package
ML040360462	List: IR 05000247/2004003
References
IR-04-003
Download: ML040360248 (23)

Text

ary 4, 2004

SUBJECT:

INDIAN POINT ENERGY CENTER UNIT 2 - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000247/2004003

Dear Mr. Dacimo:

On December 11, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Indian Point Energy Center, Unit 2. The enclosed inspection report documents the inspection findings, which were discussed on January 27, 2004, with yourself, and members of your staff.

The inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commissions rules and regulations, and with the conditions of your license. The inspection efforts included examination of selected procedures and representative records, observation of activities, and interviews with personnel.

The team concluded that, in general, problems are being properly identified, evaluated, and corrected. However, the team identified two findings of very low safety significance (Green)

involving test failures of a radiation monitor and of the technical support center battery cells.

While the equipment was determined to be functional, the team concluded that your staff did not promptly identify and address the conditions or underlying causes for the specific test failures. We consider these findings to be additional examples of the substantive cross-cutting issue in the area of problem identification and resolution, which we identified in previous assessment periods, most recently in a letter to you dated August 27, 2003. We plan to conduct an additional follow-up inspection in this area.

The team also evaluated aspects of your Design Basis Initiative (DBI) program. In August 2003, the NRC completed the Supplemental Inspection for a White finding involving a degraded fire barrier between the control room and turbine building (Inspection Report 50-247/2003-010).

At that time, the NRC concluded that Entergys corrective actions and extent-of-condition review for the specific fire barrier deficiencies were acceptable. However, the NRC also determined that additional inspection was required to confirm the adequacy of Entergy's efforts to identify and correct broader issues associated with design control. As a result, the NRC maintained the White finding open beyond the normal four quarters required by the Reactor Oversight Process, in order to complete these additional inspections. This problem identification and resolution

Mr. Fred Dacimo 2 inspection, therefore, included a review of Entergys DBI and its associated design control program. The team determined that Entergy made sufficient progress in addressing the design control issues to close the White finding. Recognizing that several multi-year DBI tasks are still in progress, the NRC will continue to monitor Entergys progress on these tasks through region-based specialists, supplemented by the strong complement of resident inspectors being maintained on-site.

In accordance with 10CFR2.790 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Wayne D. Lanning, Director Division of Reactor Safety Docket No. 50-247 License No. DPR-26

Enclosure:

NRC Inspection Report 05000247/2004003 w/Attachment: Supplemental Information

Mr. Fred Dacimo 3

REGION I==

Docket No: 50-247 License No: DPR-26 Report No: 05000247/2004003 Licensee: Entergy Nuclear Operations, Inc.

Facility: Indian Point Energy Center, Unit 2 Location: Buchanan, New York Dates: November 17-21 and December 8-11, 2003 Team Leader: B. Norris, Senior Reactor Inspector Inspectors: J. Benjamin, Reactor Inspector R. Berryman, Resident Inspector, Indian Point 3 R. Bhatia, Reactor Inspector (in-office)

G. Bowman, Reactor Inspector P. Habighorst, Senior Resident Inspector, Indian Point 2 T. Hipschman, Senior Reactor Inspector S. Iyer, Reactor Inspector T. Jackson, Project Inspector L. Scholl, Senior Reactor Inspector Observer: V. Ruuska, Observer, Finnish Radiation & Nuclear Safety Authority Approved by: Raymond K. Lorson, Chief Performance Evaluation Branch Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000247/2004003, 11/17 - 12/11/2003, Indian Point Energy Center, Unit 2; biennial baseline inspection of problem identification and resolution; problem identification and resolution.

The inspection was conducted by eight regional inspectors and two resident inspectors. Two Green findings of very low safety significance were identified. The findings were evaluated using Inspection Manual Chapter 0609, Significance Determination Process. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

Identification and Resolution of Problems The inspection team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program, evaluating and prioritizing issues, and implementing appropriate corrective actions. However, the inspectors identified two Green findings related to test failures of a radiation monitor and of the technical support center battery cells. While the equipment was determined to be functional, the team concluded that the IP2 staff did not promptly identify and address the conditions or underlying causes for the specific test failures. The inspectors considered these findings to be additional examples of the substantive cross-cutting issue in the area of problem identification and resolution identified during previous assessments. Based on interviews conducted during the inspection, station personnel felt free to identify safety issues and enter them into the corrective action program.

The team also evaluated aspects of the Design Basis Initiative (DBI) program. In August 2003, the NRC completed the Supplemental Inspection for a White finding involving a degraded fire barrier between the control room and turbine building (Inspection Report 50-247/2003-010). At that time, the NRC concluded that Entergys corrective actions and extent-of-condition review for the specific fire barrier deficiencies were acceptable. However, the NRC also determined that additional inspection was required to confirm the adequacy of Entergy's efforts to identify and correct broader issues associated with design control. As a result, the NRC maintained the White finding open in order to complete the additional inspections. This problem identification and resolution inspection, therefore, included a review of Entergy's DBI and its associated design control program. The team determined that Entergy made sufficient progress in addressing the design control issues to close the White finding.

Cornerstone: Emergency Preparedness

C

Green.

The team identified a finding of very low safety significance (Green) for the failure to properly address repetitive surveillance test failures of the R-27 plant vent noble gas effluent radiation monitor. The team determined that the licensee did not effectively identify and correct the underlying cause to preclude these repetitive test failures. After this issue was raised by the inspection team, the licensee determined that the cause of the test failures was degraded test equipment, and that the radiation monitor had been operable.

ii

The performance deficiency associated with this finding was failure to identify and address the underlying causes of repetitive failures of a TS required surveillance. The performance deficiency contributed to the monitors unavailability and subsequent test failures. The test failures of the R-27 radiation monitor adversely affected methods, systems, and equipment for assessment of radiological releases required by 10CFR50.47(b)(9). This finding was of more than minor significance because the R-27 radiation monitor was removed from service for troubleshooting periods in excess of twenty-four hours. The finding was evaluated using the Emergency Preparedness SDP, and was determined to be of very low safety significance (Green), because alternate monitoring methods were available during periods when the monitor was unavailable for troubleshooting and maintenance.

C

Green.

The team identified a finding of very low safety significance (Green) for the failure to take prompt action for out of specification indications for one cell in each of the two Technical Support Center (TSC) battery banks. While the battery banks were subsequently determined to be functional, the team concluded that the licensee did not take prompt action to either return the two battery cells to within specifications or to evaluate the acceptability of the as-found condition.

The performance deficiency associated with this finding was failure to take timely action to evaluate the degraded condition of the TSC battery cells. The degraded cells had the potential to adversely affect the facilities and equipment required to support emergency response which are required to be maintained by 10CFR50.47(b)(8). This finding was of more than minor significance because the batteries were allowed to remain in an in-determinant condition in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without adequate evaluation or compensatory measures. The finding was evaluated using the Emergency Preparedness SDP, and was determined to be of very low safety significance (Green), because the subsequent analysis indicated that the battery banks remained functional in this condition.

Licensee-Identified Violations

None.

iii

REPORT DETAILS

OTHER ACTIVITIES (OA)

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Occupational Radiation Safety, Public Radiation Safety, Physical Protection.

4OA2 Identification and Resolution of Problems (IP 71152)

a.

Effectiveness of Problem Identification

(1) Inspection Scope:

The team reviewed the procedures that described the corrective action process used by Entergy Nuclear Northeast personnel at Indian Point Unit 2 (IP2) of the Indian Point Energy Center (IPEC), and determined that problems were identified primarily through the initiation of condition reports (CRs). The team reviewed selected CRs, and attended daily management meetings where the CRs were screened for significance, to determine whether IPEC was identifying, accurately characterizing, and entering problems into the corrective action process at an appropriate threshold.

The CRs selected for review are listed in the Attachment to this report. The team chose the CRs to cover the seven cornerstones of safety identified in the NRCs Reactor Oversight Process (ROP). In addition, the team considered risk insights from IPECs probabilistic safety assessment (PSA) to focus the CR sample selection on risk significant plant equipment. The team interviewed selected plant staff to determine their understanding of the process used to address problems. Also, the team conducted walkdowns of selected areas of the plant, to independently assess whether problems were properly identified and addressed.

In addition to CRs, the team selected items from IPECs operations, maintenance, engineering, radiation protection, emergency preparedness, security, and oversight processes to verify that IPEC appropriately considered problems identified in these areas for entry into the corrective action program. Specifically, the team reviewed a sample of work orders, engineering change requests, operator log entries, control room deficiency and work-around lists, operability determinations, engineering system health reports, completed surveillance tests, installed temporary modification packages, quality assurance audit and surveillance reports, and departmental self-assessments. The documents were reviewed to ensure that underlying problems associated with each issue were appropriately considered for resolution via the corrective action process.

The documents reviewed are listed in the Attachment.

(2) Observations and Findings No findings of significance were identified.

The team concluded that IPEC personnel were generally identifying deficiencies at a low threshold, and documenting the problems on CRs, in accordance with procedure ENN-LI-102, Corrective Action Process. The CRs described and characterized the problems accurately, and, as appropriate, identified prior similar occurrences. In addition, the team noted that personnel initiated CRs for problems identified in other processes (such as work orders, engineering requests, etc.) that met the CR threshold.

The team concluded that quality assurance audits and surveillances, and department self-assessments were generally effective at identifying adverse conditions and trends.

Notwithstanding the above, during plant walkdowns, the team identified several minor equipment problems that were not entered into the corrective action process. These problems included: a small leak from the mechanical seal on #23 safety injection (SI)pump; a small oil leak from the #22 SI pump gear box; and evidence of packing leakage from a suction isolation valve (MOV-887A) for the #22 SI pump. The team discussed their findings with the system engineer and CR-IP2-2003-06956 was initiated to document these observations. The team determined that none of the above problems affected the operability of the SI system.

b.

Prioritization and Evaluation of Issues

(1) Inspection Scope:

The team reviewed the CRs listed in the Attachment to determine whether IPEC adequately evaluated and prioritized problems. The review included the appropriateness of the assigned significance, the timeliness of resolutions, and the scope and depth of the causal analysis. The CRs reviewed encompassed the full range of IPEC evaluations, including root cause analysis and apparent cause evaluations. The team selected the CRs to cover the seven cornerstones of safety identified in the ROP.

The team also considered risk insights from the PSA to focus the CR sample.

The team reviewed the CRs associated with selected non-cited violations (NCVs) to determine whether IPEC properly evaluated and resolved these issues. The team reviewed IPECs evaluation of industry operating experience information for applicability to their facility. The team also reviewed equipment operability determinations, reportability assessments, and extent of condition reviews for selected problems. The team further reviewed equipment performance results and assessments documented in completed surveillance procedures, operator log entries, and trend data to determine whether IPECs equipment performance evaluations were technically adequate to identify degrading or non-conforming equipment.

(2) Observations and Findings The team determined that generally the CRs reviewed were properly classified for significance. The team noted that significant conditions adverse to quality received a formal root cause analysis (RCA), and an extent-of-condition review. Less significant conditions adverse to quality typically received an apparent cause evaluation (ACE).

The items in the engineering and maintenance backlogs had been evaluated for risk (individually and collectively). The majority (.94%) of the CRs were for less significant issues. The level of detail provided in some of the CRs made it difficult for the inspectors to understand the issue or the resolution without additional information. The team identified two examples of inadequate evaluations that were dispositioned as Green findings.

.1 Plant Effluent Radiation Monitor (R27) Surveillance Test Failures

Introduction:

The team identified a finding of very low safety significance (Green) for the failure to properly address repetitive surveillance test failures of the R-27 plant vent noble gas effluent radiation monitor. The team determined that the licensee did not effectively identify and correct the underlying cause to preclude these repetitive test failures. After this issue was raised by the inspection team, the licensee determined that the cause of the test failures was degraded test equipment, and that the radiation monitor had been operable.

Description:

The team identified that the R-27 radiation monitor had failed five of the six quarterly surveillance tests, since July 2002. The testing was performed to demonstrate the operability of the R-27 monitor as required by Technical Specification (TS) 3.5.6.

The monitor is described in the Updated Final Safety Analysis Report (UFSAR) and is required to be maintained per NUREG-0737, Clarification of TMI Action Plan Requirements.

The R-27 radiation monitor is a single channel monitor with three detectors, one for each range (low, medium, and high). The system contains two compressors, one for the low range and the other for the medium and high ranges, which supply air samples from the plant vent to the detectors. As the level of radioactivity increases above a set value, the medium/high range compressor is designed to automatically start, and the display to automatically shift to the medium or high range, as appropriate. Surveillance test procedure (PT-Q42, Wide Range Noble Gas Monitor R-27 Functional Check)tested the response of the monitor to a simulated radiation signal. In the five surveillance test failures, the medium/high compressor did not automatically start as required.

Subsequent to each test failure, the R-27 monitor was declared inoperable and the licensee implemented the alternate sampling requirements specified in TS Table 3.5-5.

The corrective actions to restore the monitor to service included removal, inspection, and reinstallation of circuit cards; or replacement of internal components. At the time, the licensee identified some possible causes, but did not identify the underlying cause for the test failures, and did not preclude additional surveillance test failures.

When this issue was raised by the inspection team, the licensee entered the issue into the corrective action program (CR-IP2-2003-07349) and subsequently determined that the test equipment was deficient; specifically, the output from the signal generator used to develop the test signals was erratic. The team concluded that the R-27 monitor was operable during the period of the repetitive test failures.

Analysis:

The performance deficiency associated with this finding was failure to identify and address the underlying causes of repetitive failures of a TS required surveillance.

The performance deficiency contributed to the monitors unavailability and subsequent test failures. The test failures of the R-27 radiation monitor adversely affected methods, systems, and equipment for assessment of radiological releases required by 10CFR50.47(b)(9), a Risk-Significant Planning Standard described in MC-0609, Appendix B, Emergency Preparedness SDP. This finding was of more than minor significance because the R-27 radiation monitor was removed from service for troubleshooting periods in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The finding was evaluated using the Emergency Preparedness SDP, and was determined to be of very low safety significance (Green), because alternate monitoring methods were available during periods when the monitor was unavailable for troubleshooting and corrective maintenance.

Enforcement:

The team reviewed the requirements of 10 CFR 50, Appendix B and 10 CFR 50.47 and determined that this finding did not involve a violation of NRC requirements since the R-27 monitor is not safety-related and since alternate monitoring methods were available to meet the emergency plan requirements. This finding was entered into the licensees corrective action program as CR-IP2-2003-07349.

(FIN 05000247/2004003-01, Failure to Identify and Address Causes of Repetitive Surveillance Test Failures of the Plant Vent Noble Gas Effluent Monitor)

.2 Degraded Technical Support Center Batteries

Introduction:

The team identified a finding of very low safety significance (Green) for the failure to take prompt action for out of specification indications for one cell in each of the two Technical Support Center (TSC) battery banks. While the battery banks were subsequently determined to be functional, the team concluded that the licensee did not take prompt action to either return the two battery cells to within specifications or to evaluate the acceptability of the as-found condition.

Description:

The TSC batteries are the second backup electrical supply to the plant computer and the safety parameter display system computer used in the TSC to assist the control room personnel during emergency situations. The normal electrical supply for the computers is from offsite, with the TSC diesel generator being the first backup in the event of a loss of offsite electrical power.

During review of CRs-IP2-2003-06422 and -06424, the inspectors noted that, during the quarterly surveillance tests performed on October 21, 2003, one cell in each of the two TSC battery banks did not meet the acceptance criteria specified in the test procedures (TST-PT-Q-19A and B). A cell in the east bank failed for individual cell voltage (minimum acceptable value was 2.07 vdc, as-found was 2.04 vdc), and a cell in the west bank failed for specific gravity (minimum acceptable value was 1.195 specific gravity, as-found was 1.186 specific gravity). While test parameters were marginally out of specification, the team determined that the licensee did not take prompt corrective actions to either return the two indications within specifications or to evaluate the impact of the out of specification indications for the two cells on the functionality of the battery banks. The team also noted that the same battery cell in the west bank had been identified as out of specification (as-found 1.190 specific gravity) in a previous surveillance test conducted on August 1, 2003. However, the TSC battery banks were able to perform their design function during the August 14th blackout.

After the issue was raised by the inspection team, the licensee performed an evaluation and determined that each battery bank was capable of performing its required function with a single cell in each bank not meeting the acceptance criteria specified in the surveillance test and issued CR-IP2-2003-07321 to document the non-timely actions for the battery cell test failures.

Analysis:

The performance deficiency associated with this finding was failure to take timely action to evaluate the degraded condition of the TSC battery cells. The degraded cells had the potential to adversely affect the facilities and equipment required to support emergency response which are required to be maintained by 10CFR50.47(b)(8),a Non-Risk Significant Planning Standard described in MC-0609, Appendix B, Emergency Preparedness SDP. This finding was of more than minor significance because the batteries were allowed to remain in an in-determinant condition in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without adequate measures to ensure that the TSC support function would be maintained.

The finding was evaluated using the Emergency Preparedness SDP, and was determined to be of very low safety significance (Green), because the subsequent analysis indicated that the battery banks remained functional in this condition.

Enforcement:

The team reviewed the requirements of 10 CFR 50, Appendix B and 10 CFR 50.47 and determined that this finding did not involve a violation of NRC requirements since the TSC batteries are not safety-related and, since the batteries were functional, all emergency planning standards were satisfied.

(FIN 05000247/2004003-02, Failure to Evaluate the Degraded Condition of the TSC Batteries)c.

Effectiveness of Corrective Actions

(1) Inspection Scope:

The team reviewed the CRs listed in the Attachment to determine whether the actions addressed the identified causes of the problems. The team reviewed IPECs timeliness in implementing corrective actions and their effectiveness in preventing recurrence of significant conditions adverse to quality.

(2) Assessment:

No significant findings were identified in this area.

d.

Assessment of Safety Conscious Work Environment

(1) Inspection Scope:

Team members interviewed plant staff, observed various activities throughout the plant, and attended a cross section of meetings to determine if personnel were hesitant to raise safety concerns to their management and/or the NRC.

(2) Assessment:

No findings of significance were identified.

4OA5 Other Activities (IP 95001)

a.

Review of Design Basis Initiative Projects

(1) Inspection Scope The team reviewed the reconstituted design packages for three of the DBI projects against the guidelines of Entergys DBI Project Plan: BR-2, Condition Reports; DB-3, Test Design Basis Review; and PI-4, Hydraulic Modeling. The team also reviewed the IP2 Electrical Distribution System Load Flow

Analysis.

In addition, the team reviewed the self-assessments of the completed DBI projects: BR-3, Work Orders on Engineering Hold; DB-5, Heatup and Cooldown Curves; PI-4, Hydraulic Modeling; PI-5, ISI/IST Quality Group Classification and Boundaries; WIRE-2, Gas Turbine Wiring Verification; and the High Energy Line Break (HELB) Basis Reconstitution.

(2) Observations and Findings No findings of significance were identified relative to the quality of the reviewed DBI project packages or the status of the ongoing projects.

The BR-2 project was ongoing at the time of the inspection with 35 of the original 51 condition reports (CRs) remaining open. The CRs were to be maintained open until all of the associated corrective actions were completed. The DB-3 project was developed to ensure that procedural revisions resulting from the TS reviews were tracked through completion. The primary objective of the PI-4 project was completed, involving the development of hydraulic models for selected systems; however, a second objective on the PI-4 project plan, involving the development of a method to maintain the models current and to control their use, had not been completed. The inspectors verified that the electrical load flow analysis met the design basis requirements during normal and abnormal operating and shutdown conditions.

The licensee performed self-assessments of the completed individual DBI project packages. The team determined that the self-assessments were generally critical but identified one minor issue where a CR was not initiated for an observation related to the completion of the PI-4 project plan. Entergy subsequently initiated a CR for this oversight (CR-IP2-2003-06994). The team identified some minor observations related to updating of the DBI project plan. The team discussed these observations with the DBI Project Manager.

.1 (Closed) URI 05000247/2003004-02: Lack of Basis for Functionality of Backup CCW

Water Sources During an engineering design inspection (NRC IR 50-247/2003-004, March 2003), the inspectors identified a lack of an engineering calculation or testing to support that primary water and city water were capable of providing backup cooling for CCW heat loads, as described in the UFSAR. The team reviewed an engineering analysis completed in May 2003, and discussed it with cognizant personnel. Using the existing CCW pipe flow model for the assessment, the licensee performed an analysis which demonstrated that makeup from city water and primary water could provide adequate backup cooling to the SI, residual heat removal, and charging pump coolers. The team determined that this assessment was reasonable and also reviewed the capacity of the existing floor drains in the SI, residual heat removal, and charging pump rooms to ensure that the discharged city water cooling flow could be evacuated from the rooms.

The team confirmed that the floor drains could accommodate the flow rates from the coolers, as referenced by the internal flooding analysis.

The team determined that the failure to develop an adequate design calculation to support the UFSAR assumptions regarding the availability of backup cooling for the CCW heat loads was a violation of 10CFR50, Appendix B, Criterion III (Design Control).

However, the team determined that this issue was of minor significance and not subject to enforcement action in accordance with Section IV of the NRCs Enforcement Policy.

This unresolved item is closed.

.2 (Closed) URI 05000247/2003004-03: Lack of Basis for CCW Flow Requirements for the

Recirculation and Safety Injection Pumps During an engineering design inspection (NRC IR 50-247/03-004, March 2003), the inspectors identified a lack of an engineering basis for minimum CCW flow for SI recirculating pump motor coolers and SI pump lube oil coolers during design basis conditions. The team reviewed and discussed with cognizant personnel the SI pump oil cooler design flow rates. Entergy engineering calculation PGI-0186-00 concluded that a design flow of 1.9 gallons per minute to the lube oil coolers was necessary during design basis conditions. At the time of the inspection, CR-IP2-2003-00912 contained an open corrective action to revise plant documents to reflect this design flowrate. Past surveillance test results provided adequate assurance that the revised design flow rates have been maintained.

The team determined that the failure to develop an adequate engineering basis for the minimum CCW flow to SI components during design basis conditions was a violation of 10CFR50, Appendix B, Criterion III (Design Control). However, the team determined that this issue was of minor significance and not subject to enforcement action in accordance with Section IV of the NRCs Enforcement Policy. This unresolved item is closed.

.3 (Closed) URI 05000247/2003004-04: Lack of Calculation for Battery Sizing to Support

the Alternate Offsite Power and ASSS Circuit Breaker Operation During an engineering design inspection (NRC IR 50-247/03-004, March 2003), the inspectors identified a lack of a sizing and load calculation for the Unit 1 DC battery system. The Unit 1 batteries support the control and protection circuits for the 13.8 kV and 440 volt circuit breakers used to provide alternate power during various fire safe shutdown scenarios.

Entergy completed calculation FEX-00201-00, IP1 Voltage Profiles for Battery 11 and 12 Demonstrated that Alternate Safe Shutdown Electrical Buses 12RW3, 12FD3, and 13.8 kV Lighting and Power Bus Section 3. The calculation concluded that adequate battery voltage existed to support breaker operation. The team reviewed the calculation and identified that the batteries provided the required voltage except for one minute on battery 11 (107 vdc) and one minute on battery 12 (110 vdc). The licensee did not provide a basis for why this was acceptable, however, following subsequent review, Entergy provided additional information related to design assumptions and the team concluded that the minimum terminal voltage would be satisfied.

The licensees documented basis for this condition also included that breaker manipulations could be considered manual for fire scenarios involving safe shutdown.

License condition 2.K states ... the alternate safe shutdown system components powered from Indian Point Unit 1 switchgear do not rely on component power or control power from any IP2 buses when transferred to IP1 power supply by transfer switches.

The licensee will develop and implement written procedures for obtaining safe shutdown conditions given a fire event. The team reviewed Entergys abnormal operating instruction (AOI) 27.1.9, Control Room Inaccessibility Safe Shutdown Control, and determined that it did not provide guidance to the operators to manually close the 440 volt breakers or the 13.8 kV breakers. The inspector also confirmed through discussions with operators that routine training was not provided for manually closure of 440 volt or 13.8 kV breakers.

The team determined that the failure to develop adequate procedures for manual operation of the breakers was a violation of License Condition 2.K. However, the team determined that this issue was of minor significance and not subject to formal enforcement action in accordance with Section IV of the NRCs Enforcement Policy.

This unresolved item is closed.

.4 (Closed) VIO 50-247/02-010-001 (White): Violation of License Condition 2.K and the

Approved Fire Protection Program Involving the Failure to Implement and Maintain a Rated Three-hour Fire Barrier Between the Control Room West Wall and the Turbine Building.

The team reviewed elements of Entergys ongoing DBI and associated design control program to confirm the adequacy of Entergys efforts to identify and correct the broad issues which contributed to the White finding involving a degraded three-hour fire barrier between the control room and turbine building (NRC Inspection Report 50-247/2002-010, dated August 26, 2002). A previous supplemental inspection (Inspection Report 50-247/2003-010, dated August 4, 2003) determined that Entergys corrective actions and extent-of-condition review for the specific fire barrier deficiencies were acceptable.

However, the NRC determined that additional inspection was required to review progress of the DBI. The NRC determined that, while the DBI effort has additional multi-year tasks to complete, adequate progress has been made to allow continued NRC review of the design control program through the baseline inspection program focusing on engineering and corrective action effectiveness. Based upon the results of this inspection, and the supplemental inspection, the White finding is closed.

4OA6 Meetings, including Exit

On December 11, 2003, the team conducted a de-brief of the preliminary inspection with Mr. C. Schwarz, General Manager - Plant Operations, and other members of the IPEC staff. The inspectors confirmed that no proprietary information was being retained.

On January 27, 2004, the NRC conducted a telephone exit meeting with Mr. F. Dacimo, Site Vice President, and other members of the IPEC staff, at which time the final inspection results were presented.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Barry Superintendent, Plant Security

W

Blair Manager, Licensing

.

C. Brown Supervisor, Maintenance Testing

J. Comiotes Director - Nuclear Safety Assurance

J. Deroy General Manager - Engineering

J. Donnelly Manager, Corrective Action

K. Finucan Emergency Preparedness Engineer

D. Gately Assistant Radiation Protection Supervisor

P. Gropp Project Manager, Design Basis Initiative

J. Hill Engineering Supervisor

J. Janicki Supervisor, Operations Procedures

T. Jones Licensing Engineer

F. Marcussen Manager, Security Operations

T. McCaffrey Manager, System Engineering

J. McCann Manager, Corporate Licensing

J. Perotta Manager, Quality Assurance

S. Petrosi Manager, Design Engineering

J. Raffaele Supervisor, Electrical Design Engineering

J. Reynolds Supervisor, Corrective Action

C. Schawrz General Manager - Plant Operations

B. Taggart Employee Concerns Coordinator

J. Ventosa Manager, Operations

A. Williams Manager, Unit 2 Operations

ITEMS OPENED, CLOSED, AND UPDATED

Opened and Closed:

FIN 05000247/2004003-01 Failure to Identify and Address Causes of Repetitive Surveillance

Test Failures of the Plant Vent Noble Gas Effluent Monitor

(Section 4OA2.b(2).1)

FIN 05000247/2004003-02 Failure to Evaluate the Degraded Condition of the TSC Batteries

(Section 4OA2.b(2).2)

Closed:

URI 05000247/2003004-02 Lack of Basis for Functionality of Backup CCW Water Sources

(Section 4OA5.b(2).1)

URI 05000247/2003004-03 Lack of Basis for CCW Flow Requirements for the Recirculation

and Safety Injection Pumps (Section 4OA5.b(2).2)

URI 05000247/2003004-04 Lack of Calculation for Battery Sizing to Support the Alternate

Offsite Power and ASSS Circuit Breaker Operation

(Section 4OA5.b(2).3)

VIO 05000247/2002010-01 Violation of License Condition 2.K and the Fire Protection

Program Involving the Failure to Implement and Maintain a

Three-hour Barrier Between the Control Room West Wall and

the Turbine Building (Section 4OA5.b(2).4)