IR 05000313/2004010

IR 05000313-04-010, on 09/27/2004 - 10/29/2004; Entergy Operations, Inc., Arkansas Nuclear One, Units 1 and 2; Triennial Fire Protection Inspection

ML050310098
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/28/2005
From:	Laura Smith NRC/RGN-IV/DRS/EMB
To:	Forbes J Entergy Operations
References
IR-04-010
Download: ML050310098 (38)

Text

ary 28, 2005

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 and 2 - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000313/2004010; 05000368/2004010

Dear Mr. Forbes:

On September 27, 2004 through October 29, 2004, the NRC conducted a triennial fire protection inspection at Arkansas Nuclear One, Units 1 and 2. Additional inspection activities continued through December 14, 2004. The enclosed report documents the inspection findings which were discussed on December 14, 2004, with Mr. Clifford Eubanks, General Manager, Plant Operations and other members of your staff.

During this triennial fire protection inspection, the inspection team examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and the conditions of your license. The inspection team visually inspected selected fire zones, interviewed operators and fire protection staff, reviewed selected procedures and records, and stepped-through operator actions prescribed in selected fire protection procedures.

Based on the results of this inspection, no findings of significance were identified.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

//RA//

Linda Joy Smith, Branch Chief Plant Engineering Branch Division of Reactor Safety

Entergy Operations, Inc. -2-Dockets: 50-313 50-368 Licenses: DPR-51 NPF-6

Enclosures:

Inspection Report 05000313/2004010; 05000368/2004010 w/Attachments:

1. Supplemental Information 2. ANO Position on the Requirements of 10 CFR Part 50, Appendix R, Section III.L 3. Summary of Integrated Calculation Process Used at ANO for App. R 4. Integrated Analysis Discussion 5. Evaluation of Pressurizer Level

REGION IV==

Docket(s): 50-313; 50-368 License(s): DPR-51; NPF-6 Report No.: 05000313/2004010; 05000368/2004010 Licensee: Entergy Operations, Inc.

Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Hwy. 64W and Hwy. 333 South Russellville, Arkansas Dates: September 27, 2004 through October 29, 2004 Team Leader R. L. Nease, Senior Reactor Inspector Plant Engineering Branch Inspectors: G. Replogle, Senior Reactor Inspector Plant Engineering Branch Paula Goldberg, Reactor Inspector Plant Engineering Branch Tim McConnell, Reactor Inspector Plant Engineering Branch Accompanying Dean Overland, Reactor Inspector Personnel Plant Engineering Branch Tony Brown, Technical Support Staff Contractor Kenneth Sullivan, Project Engineer Brookhaven National Laboratory Approved By: Linda Joy Smith, Chief Plant Engineering Branch Enclosure

SUMMARY OF FINDINGS

IR 05000313/2004-010; 05000368/2004-010; September 27, 2004 through October 29, 2004;

Arkansas Nuclear One, Units 1 and 2: Arkansas Nuclear One, Units 1 and 2; Triennial Fire Protection Inspection This report covered an announced inspection by four region-based inspectors, two accompanying personnel from NRC Region IV, and one contractor. One unresolved item (URI)was identified. The significance of most findings is indicated by its color (Green, White, Yellow,

Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

NRC-Identified

Finding None B. Licensee-Identified Findings None

REPORT DETAILS

1R05 Fire Protection

The purpose of this inspection was to review the Arkansas Nuclear One (ANO) facility fire protection program for selected risk-significant fire areas. Emphasis was placed on verification of the licensee's post-fire safe shutdown capability. The inspection team performed this inspection using the guidance in Inspection Procedure 71111.05, Fire Protection, which requires selecting three to five fire areas for review. The inspection was performed in accordance with the NRC regulatory oversight process using a risk-informed approach for selecting the fire areas and attributes to be inspected. The team used licensee Calculation 85-E-0053-47, Individual Plant Examination of External Event/Fire, to choose several risk-significant areas for detailed inspection and review. The following three fire zones were chosen for review during this inspection:

• Fire Zone 2109U (Corridor); located in Fire Area JJ, Unit 2

• Fire Zone 2097X (East DC Equipment Room); located in Fire Area SS, Unit 2

• Fire Zone 2091BB (Electrical Equipment Area); located in Fire Area B, Unit 2 For each of these fire areas, the inspection focused on fire protection features, systems and equipment necessary to achieve and maintain safe shutdown conditions, and licensing basis commitments.

In accordance with NRC Inspection Procedure 71111.05, dated March 6, 2003, the inspection did not include a comprehensive review of the potential impact of fire-induced failures in associated circuits of concern to post-fire safe shutdown. In response to a March 2001 voluntary industry initiative, the scope of NRC Inspection Procedure 71111.05 has been temporarily reduced pending the resolution of specific review criteria for fire-induced circuit failures of associated circuits.

Documents reviewed by the team are listed in the attachment to this inspection report.

.1 Systems Required to Achieve and Maintain Post-Fire Safe Shutdown

a. Inspection Scope

For the selected fire zones, the team reviewed the licensee's methodology for achieving and maintaining post fire safe shutdown described in Calculation 85-E-0086-01, "Safe Shutdown Capability Assessment Unit 1," and Calculation 85-E-0087-01, "Safe Shutdown Capability Assessment, Unit 2." This review was performed in order to ensure that at least one post-fire safe shutdown success path was available in the event of a fire in each of the selected areas. In addition the team verified that the licensee had properly identified the systems and component required to achieve and maintain safe shutdown conditions. The team focused on the below-listed functions that must be available to achieve and maintain post-fire safe shutdown conditions. In addition, the team verified that process monitoring capable of providing direct readings to perform and control these functions was available.

• Reactivity control capable of achieving and maintaining cold shutdown reactivity conditions,

• Reactor coolant makeup capable of maintaining the reactor coolant inventory,

• Reactor heat removal capable of achieving and maintaining decay heat removal, and

• Supporting systems capable of providing all other services necessary to permit extended operation of equipment necessary to achieve and maintain hot shutdown conditions.

To assure the licensee had properly identified the components and equipment necessary to achieve and maintain safe shutdown conditions in the event of a fire in the fire areas selected for review, the team reviewed piping and instrumentation diagrams for the systems required for performing above-listed functional requirements, and compared them to the list of equipment documented in the licensees post-fire safe shutdown analysis. In addition, plant drawings, operating procedures, operator lesson plans, and other relevant documents were reviewed to verify the flow paths and operational characteristics of those systems relied on to accomplish the 10 CFR Part 50, Appendix R post-fire safe shutdown functions listed above.

b. Findings

No findings of significance were identified.

.2 Fire Protection of Safe Shutdown Capability

a. Inspection Scope

For each of the selected fire areas, the team reviewed licensee documentation to verify that at least one train of equipment needed to achieve and maintain hot shutdown conditions was free of fire damage in the event of a fire in the selected fire areas.

Specifically, the team examined (on a sampling basis) the separation of safe shutdown cables equipment and components within the same fire areas to verify that the licensee met the requirements of 10 CFR Part 50, Appendix R, Section III.G.2 The team reviewed the licensee's methodology for meeting the requirements of 10 CFR 50.48, 10 CFR Part 50, Appendix R, and the bases for the NRC's acceptance of this methodology as documented in NRC safety evaluation reports. In addition, the team reviewed license documentation, such as, the Arkansas Nuclear One, Units 1 and 2 Safety Evaluation Reports, submittals made to the NRC by the licensee in support of the NRC's review of their fire protection program, and exemptions from NRC regulations to verify that the licensee met license commitments.

b. Findings

Fire Zones 98J (Unit 1 diesel generator corridor) and 99M (Unit 1 north switchgear room) were the subject of a finding of low to moderate significance (white) and a Notice of Violation issued by letter dated April 7, 2004. This finding involved the failure to ensure that on train of required safe shutdown equipment (including cables) was free of fire damage in accordance with 10 CFR Part 50, Appendix R, Section III.G.2. During their extent of condition evaluation, the licensee identified that the team's selected fire zones, Fire Zones 2109U (Unit 2 corridor); 2097X (Unit 2 east DC equipment room); and 2091BB (Unit 2 electrical equipment area) were also subject to this finding.

For the fire zones selected for review, no additional findings of significance involving the separation and protection requirements of 10 CFR Part 50, Section III.G.2 were identified by the team during this inspection.

.3 Post-fire Safe Shutdown Circuit Analysis

a. Inspection Scope

On a sample basis, the team verified that cables of equipment required to achieve and maintain hot shutdown conditions in the event of fire in selected fire zones had been properly identified and either adequately protected from the potentially adverse effects of fire damage or analyzed to show that fire-induced faults (e.g., hot shorts, open circuits, and shorts to ground) would not prevent safe shutdown. Cable routing data depicting the routing of power and control cables associated with each of the selected components was reviewed. The specific components selected for review are listed below.

Electrical distribution components:

DC Panel 2D23 DC Panel 2D24 2D24 Inverter Y11 Inverter Y22 Reactor coolant system inventory makeup components:

2CV-4873-1 (charging pump suction valve from volume control tank)2CV-4950-2 (charging pump suction valve from reactor water storage tank)2CV-4824-2 (auxiliary spray valve)2P36A, 2P36B and 2P36C 2P36C (charging pumps)

Potential reactor coolant system leak path components:

2CV4698-1 and 2CV4740-2 (pressurizer vent valves)2CV-4823-2, 2CV-4821-1and 2CV-4820-2 (letdown valves):

Decay heat removal components:

2CV-1001, 2CV-1002, 2CV-1051 and 2CV-1052 (atmospheric dump valves)2P7A and 2P7B (emergency feedwater pumps)

In addition, on a sampling basis, the team reviewed the adequacy of selected electrical protective devices (e.g., circuit breakers, fuses, relays), breaker coordination, and the adequacy of electrical protection provided for nonessential cables, which share a common enclosure (e.g., raceway, junction box, conduit, etc) with cables of equipment required to achieve and maintain safe shutdown conditions.

For the selected fire areas, the team also reviewed the location and installation of diagnostic instrumentation that is necessary for achieving and maintaining safe shutdown conditions to ensure that in the event of a fire, this instrumentation would remain functional.

b. Findings

No findings of significance were identified.

.4 Alternative Safe Shutdown Capability

a. Inspection Scope

The team reviewed the licensee's alternative shutdown methodology to determine if the licensee properly identified the components and systems necessary to achieve and maintain safe shutdown conditions from alternative shutdown locations in the event of a fire in the control room, requiring control room evacuation. The team focused on the adequacy of the systems selected for reactivity control, reactor coolant makeup, reactor heat removal, process monitoring and support system functions. The team verified that the licensee's methodology included an evaluation that hot and cold shutdown from outside the control room can be achieved and maintained with off-site power available or not available. The team verified that the transfer of control from the control room to the alternative locations was not affected by fire-induced circuit faults by reviewing the provision of separate fuses for alternative shutdown control circuits. The team also reviewed plant technical specifications and applicable surveillance procedures to verify incorporation of operability testing of alternative shutdown instrumentation and transfer of control functions.

b. Findings

Introduction:

The team identified an unresolved item (URI) concerning an inadequate alternate shutdown analysis. The alternate shutdown analysis was inadequate because:

(1) it was based on acceptance criteria, which was inconsistent with 10 CFR 50, Appendix R requirements; and

(2) did not accurately model the simultaneous operation of emergency feedwater restoration, letdown isolation and makeup restoration.

Description:

The licensee developed Calculation 85-E-0086-02, "Manual Action Feasibility Methodology and Common Results," Revision 0, to demonstrate alternative shutdown capability for ANO, Units 1 and 2. This calculation did not directly address alternate shutdown, but was instead, a compilation of other calculations. The team identified that

(1) these analyses were based on acceptance criteria that differed from that in 10 CFR Appendix R,Section III.L; and

(2) the licensee had not performed an integrated analysis which considered the results of more than one set of analyzed conditions occurring simultaneously. Each of these issues is discussed below.

Inadequate Alternative Shutdown Acceptance Criteria: Title 10 of the Federal Code of Regulations, Part 50, Appendix R,Section III.L. 1 specifies, in part, that during alternative post-fire shutdown, the reactor coolant system process variables (reactor temperature, pressure, and level) shall be maintained within those predicted for a loss of normal AC power. The licensee did not have predictive calculations for a loss of normal AC power, but utilized other existing calculations to predict plant response to this event.

The team reviewed the following calculations, and noted that these calculations predicted that reactor coolant level would remain well within the level indication in the pressurizer.

• (Unit 1) Calculation BWNP-20007, Number 86-1118045-00 (steam generator tube rupture and loss of normal ac power)

• (Unit 1) Framatome Technologies Calculation 32-1266115-00 (ANO-1 turbine trip analysis)

• (Unit 2) Calculation ANO-2 95-E-0080-05 (ANO-2 loss of condenser vacuum)

Title 10 of the Federal Code of Regulations, Part 50, Appendix R,Section III.L.2 specifies in part, that the reactor coolant makeup function shall be capable of maintaining the reactor coolant level within the level indication in the pressurizer. The team reviewed a sampling (listed below) of the calculations that comprised Calculation 85-E-0086-02, "Manual Action Feasibility Methodology and Common Results," which the licensee used to demonstrate alternative shutdown capability. The team identified that the acceptance criteria were not consistent with maintaining the reactor coolant level within the level indication in the pressurizer (III.L.2) nor were they consistent with maintaining reactor coolant level within that predicted for a loss of normal AC power (III.L.1). Furthermore, the acceptance criteria among the various calculations were not consistent with each other

• Unit 1, Calculation 89-E-0047-20, Time to Restore Emergency Feedwater, Revision 1 (calculated the time to reach to top of the fuel)

• Unit 1, Calculation 85-E-0072-03, Time to Establish Positive Control of RCS Inventory, (estimated the amount of time for level to reach the pressurizer surge line)

• Unit 1, Calculation 1CNAA039401, Time to Isolate Main Feedwater, (found the loss of indicated pressurizer level during the event to be acceptable).

• Unit 2, Calculation 85-E-0072-04, Normal/Excess Letdown Inventory Loss (determined the time to empty the pressurizer surge line)

• Unit 2, Calculation 87-E-0003-01, Time to Isolate Main Feedwater (determined the time necessary to experience RCS loop voiding).

The inspectors did not perform an exhaustive search of all possible supporting calculations that could be affected. The licensee acknowledged that, in general, these calculations did not use acceptance criteria consistent with those in 10 CFR Part 50, Appendix R, Section III.L. However, the licensee disagreed that maintaining level within the pressurizer indicating range was a requirement of 10 CFR Part 50, Appendix R, Section III.L.2. The licensee also disagreed that maintaining level within that predicted by a normal loss of AC power was a requirement of 10 CFR Part 50, Appendix R, Section III.L.1. This licensee submitted an analysis, entitled "ANO Position on the Requirements of 10 CFR Part 50, Appendix R, Section III.L" describing their position, which is Attachment 2 to this report.

Failure to Integrate Analyses: The team identified that the licensee had not performed an integrated analysis, which considered the results of more than one set of analyzed conditions occurring simultaneously. For example, the Time to Restore Emergency Feedwater, analysis was developed independently of the Time to Establish Positive Control of RCS Inventory, analysis. For alternate shutdown purposes, both events could occur simultaneously and the results of one analysis could impact the results of the other in a nonconservative direction. The licensee acknowledged that they did not have an integrated alternative shutdown analysis, merely a compilation of existing calculations performed for other events. They posted a fire impairment for the control rooms and initiated Condition Reports CR-ANO-C-2004-0755 and CR-ANO-C-2004-1758 to address the lack of a comprehensive alternative shutdown analysis, and to develop an analysis specific to fire safe shutdown. The licensee provided a paper summarizing their integrated calculation process entitled, "Summary of the Integrated Calculation Process Used at ANO for App. R," (Attachment 3 to this report) to demonstrate that their alternative shutdown methodology was adequate. The licensee also submitted a discussion paper on their integrated calculation process entitled, "Integrated Analysis Discussion," (Attachment 4 to this report). In this paper the licensee concluded that their approach is conservative and suitable for use until an analysis specific to 10 CFR Part 50, Appendix R can be developed.

The use of inappropriate and inconsistent acceptance criteria and the failure to perform an integrated analysis was a concern, because the licensee based the times that operators needed to perform actions required to achieving and maintaining safe shutdown on the time limits generated by these calculations. As stated above, the licensee posted a fire impairment and entered this issue into their corrective action program Condition Reports CR-ANO-C-2004-0755 and CR-ANO-C-2004-1758.

Analysis:

This issue is unresolved pending further NRC review of its compliance aspects. It's significance will be determined upon resolution of the URI.

Enforcement:

Title 10 of the Federal Code of Regulations, Part 50, Appendix R,Section III.L. 1 specifies, in part, that during alternative post-fire shutdown, the reactor coolant system process variables (reactor temperature, pressure, and level) shall be maintained within those predicted for a loss of normal AC power. Title 10 of the Federal Code of Regulations, Part 50, Appendix R,Section III.L.2 specifies in part, that the reactor coolant makeup function shall be capable of maintaining the reactor coolant level within the level indication in the pressurizer. The licensee failed to demonstrate by analysis that they had implemented an alternative shutdown methodology that (1)maintained reactor coolant process variables (e.g., reactor level) within those predicted for a normal loss of AC power (Appendix R,Section III.L.1); and

(2) maintained reactor coolant level within the level indication in the pressurizer (Appendix R,Section III.L.2).

The licensee submitted a paper (attached) contesting the team's position regarding the requirements of 10 CFR Part 50, Appendix R, Sections III.L.1 and III.L.2. This issue is unresolved pending further NRC review of the compliance aspects (URI 05000313; 368/2004010-01, Failure to Maintain Reactor Inventory Within the Pressurizer Indicating Range and Inadequate Alternate Shutdown Procedure).

.5 Operational Implementation of Alternative Safe Shutdown

a. Inspection Scope

The team reviewed the systems required to achieve alternative safe shutdown to determine if the licensee had properly identified the components and systems necessary to achieve and maintain safe shutdown conditions from the remote shutdown panel.

The team focused on the adequacy of the systems to perform reactor pressure control, reactor makeup, decay heat removal, process monitoring, and support system functions.

The team reviewed Procedures 1203.002, "Alternate Shutdown," (Unit 1) and 2203.014, "Alternate Shutdown," (Unit 2), which would be used by operators to shut down the reactor in the event of a fire requiring evacuation control room evacuation. The team also walked through the procedures with licensed operators to determine its adequacy to direct actions necessary to achieve and maintain safe shutdown conditions in accordance with their safe shutdown analysis.

b. Findings

Introduction.

The team identified an issue involving the failure to implement a fire protection procedure. Specifically, in walking through Procedure 1203.002, "Alternate Shutdown," the team noted that licensed operators would not have been able to complete certain portions of the procedure within the times assumed in Calculation 85-E-0086-02, "Manual Action Feasibility Methodology and Common Results," possibly resulting in reactor level falling below pressurizer indication. The failure to perform a rigorous validation of the procedure contributed to this issue. As discussed above (in Section 1R05.4 of this report) the team considered that in implementing a methodology based on reactor level falling below pressurizer indication the licensee is not meeting 10 CFR Part 50, Sections III.L.1 and III.L.2. This issue is unresolved pending further NRC review of the compliance aspects.

Discussion. The team performed a walkdown of Procedure 1203.002, Alternate Shutdown, (Unit 1), with plant operators to verify that operator actions could be completed within times specified in their post fire safe shutdown analyses. As stated in above, Calculation 85-E-0086-02, "Manual Action Feasibility Methodology and Common Results," is a compilation of several existing calculations performed for other events.

(Note: the adequacy of this calculation is discussed above in Section 1R05.4 of this report). The team found that the operators could not perform some of the required actions within the times specified in the calculations. For example, Calculation 85-0072-03, Time To Loss of Subcooling or Loss of Pressurizer Liquid Inventory From Plant Trip with no Makeup Available Under Various Leak Path Scenarios, specified, in part, that the time before normal makeup flow must be established to maintain level within the pressurizer (actually empty the surge line) is:

With AC power - letdown isolated in 2.9 minutes, makeup initiated in 30 minutes With RCPs - Letdown isolated in 5 minutes, makeup initiated in 5 minutes No AC power - Letdown isolated in 6.9 minutes, makeup initiated in 30 minutes Letdown isolated in 8 minutes, makeup initiated in 8 minutes Restore makeup in 55 minutes, based on letdown isolation in about 2 minutes for a realistic letdown flow rate.

During the walkdown of Procedure 1203.002, the team noted that operators would have isolated letdown in approximately 5 minutes and initiated makeup in approximately 55 minutes with RCP secured and no AC power available. These operator action times do not meet any of the above calculated times to secure letdown and establish make-up.

The team identified that a lack of rigorous procedure validation contributed to this issue.

Procedure writers had informally estimated the times necessary to perform procedural steps, rather than time the actions in the plant. The inspectors identified the following problems with this informal validation effort:

• The time estimates were sometimes not credible. For example some steps were estimated to be completed in 2 seconds or less, including reading the step(s),transit time, identifying components, turning switches and self checking.

• The licensee did not consistently account for the times necessary to manually operate motor-operated valves.

• The time estimate for ensuring that both emergency diesel generators were operating was only two minutes (total for both diesels). This could entail starting the diesels if they weren't operating. However, the team observed that it took operators approximately 12 minutes to manually start each unit. The procedure writers had estimated only the time to verify that each diesel generator was operating, but had not estimated the time it would take to start the diesel generators.

The licensee provided an analysis discussing the consequences of operators not securing letdown and establishing make-up within the analyzed timelines. This analysis entitled, "Evaluation of Pressurizer Level," is provided as Attachment 5 to this report. In this analysis, the licensee concluded that the time for restoring the make-up function would be extended, as long as operators tripped the reactor coolant pumps prior to level going below the bottom of the pressurizer surge line. The team determined that permitting pressurizer level to drop below the indicating range does not meet the requirements of 10 CFR Part 50, Sections III.L.1 and III.L.2. As discussed above in Section 1R05.4 of this report, the licensee challenged the team's view of these requirements. The team considered the resolution of this issue to be dependent on the resolution of URI 05000313; 368/2004010-01, discussed above in Section 1R05.4 of this report. Therefore this issue is part of URI 05000313; 368/2004010-01 which is unresolved pending further NRC review of its compliance aspects.

The licensee initiated a fire impairment and entered this finding into their corrective action program as Condition Report CR-ANO-C-2004-01758

Analysis:

The issue is greater than minor because it has the potential to impact the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (such as fire) to prevent undesirable consequences. The consequence of implementing an alternate shutdown procedure in which operators could not secure letdown and establish make-up in accordance with an analyzed timeline would be the potential for reactor coolant level to fall below pressurizer indication. This issue is part of URI 05000313; 368/2004010-01, which is unresolved pending further review of its compliance aspects. The significance of this issue will be determined resolution of this URI.

Enforcement.

Technical Specification 5.4.1.c requires the licensee, in part, to establish and maintain procedures covering fire protection program implementation. Contrary to the above, the licensee implemented a fire protection alternate shutdown procedure (Procedure 1203.002), in which operators could not perform some of their actions within the times analyzed in their alternative shutdown analysis. The consequence of not meeting the analyzed time line would be the potential for reactor level falling below the indicating range of the pressurizer. The requirement in 10 CFR Part 50, Appendix R, Section III.L.2 to maintain level within the indicating range of the pressurizer has been challenged by the licensee, as discussed above in Section 1R05.4 of this report. This issue is part of URI 05000313; 368/2004010-01, Failure to Maintain Reactor Inventory Within the Pressurizer Indicating Range and Inadequate Alternate Shutdown Procedure), which is unresolved pending further review of its compliance aspects.

.6 Communications

a. Inspection Scope

The team reviewed the communication systems required to implement fire fighting and operations to achieve and maintain a safe shutdown condition. The team reviewed the plant radio system and the PAX (telephone) system which were to be used by operations personnel to perform an alternative shutdown outside of the control room.

The team reviewed the design of the radio system to

(1) ensure the radio system was sufficient to support alternative shutdown operator actions, and

(2) ensure that damage from a control room fire will not impact the performance of the rest of the system. The team also reviewed the use of the portable radio system for use during fire fighting activities. The portable communication systems were reviewed for the impact that damage from fires in the selected fire areas could have on the licensee's ability to achieve and maintain safe shutdown conditions. This review included verification that the design of the systems was adequate to support operator and fire brigade actions, as applicable.

b. Findings

No findings of significance were identified.

.7 Emergency Lighting

a. Inspection Scope

The team reviewed the adequacy of emergency lighting for performing actions required in Procedures 2203.014, "Alternate Shutdown Unit 2, " Change 015-03-0, and 1203.002, "Alternate Shutdown Unit 1, Change 015-06-0, which included access and egress routes. The team reviewed test procedures, test data, and battery trending to verify that the individual battery operated units were able to supply light for the required 8-hour period. The team also reviewed emergency light drawings.

b. Findings

No findings of significance were identified.

.8 Cold Shutdown Repairs

a. Inspection Scope

The team reviewed the licensee's safe shutdown circuit analysis and plant procedures for responding to fires and implementing safe shutdown activities in order to determine if any repairs were required in order to achieve cold shutdown. The licensee had identified two systems (Decay Heat Removal and Low Pressure Safety Injection) that could potentially require repair. The repairs to these systems consisted of the replacement of parts to three control valves (CV-1428, CV-1429, and CV-5017-1) that could be overtorqued in the event of a fire. The replacement of these damaged parts would restore the ability to reach cold shutdown based on the safe shutdown methodology implemented. The team verified that the replacement parts and tools were available and the procedure to perform the repairs was feasible. The team also evaluated whether cold shutdown could be achieved within the required time using the licensee's procedures and repair methods.

b. Findings

No findings of significance were identified.

.9 Compensatory Measures

a. Inspection Scope

The team verified, by sampling, that adequate compensatory measures were put in place by the licensee for out-of-service, degraded, or inoperable fire protection features and post-fire safe shutdown equipment, and systems. The team reviewed the items on the fire impairment list in effect at the time of the inspection and compared them to the fire areas receiving hourly fire watch rounds. The team reviewed the fire protection impairment list to verify that the impairments had been entered into the licensees corrective action program and that corrective actions to restore the impaired equipment were timely and appropriate.

b. Findings

No findings of significance were identified.

.10 Fire Protection Systems, Features, and Equipment

a. Inspection Scope

For the selected fire areas, the team evaluated the adequacy of selected fire protection features, such as fire suppression and detection systems, fire area barriers, penetration seals, and fire doors. The team observed the material condition and configuration of the installed fire detection and suppression systems, fire barriers, and construction details and supporting fire tests for the installed fire barriers. In addition, the team reviewed license documentation, such as NRC safety evaluation reports and deviations from NRC regulations and the National Fire Protection Association codes to verify that fire protection features met license commitments.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

On October 29, 2004, the team leader presented preliminary inspection results to Mr.

Cliff Eubanks, General Manager of Plant Operations and other members of his staff who acknowledged the findings. On December 14, 2004, the team leader presented the final inspection results in an exit meeting to Mr. Clifford Eubanks, General Manager, Plant Operations, and other members of the licensee's staff , who acknowledged the findings.

The team leader confirmed that proprietary information was not provided or examined during this inspection.

ATTACHMENTS:

4. Integrated Analysis Discussion

ATTACHMENT 1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Dukes, Consultant, NISYS Corporation

C. Eubanks, General Manager, Plant Operations

J. Forbes, Vice President, Operations

B. Greeson, Acting Engineering Programs and Components Manager

R. Hendrix, Fire Protection Technical Specialist

D. James, Manager, Licensing

J. Johnson, Fire Protection Specialist

E. Kleinsorg, Consultant, Kleinsorg Group

J. Kowalewski, Director, Engineering

R. Loveland, Reactor Operator

K. Parkinson, Consultant

R. Puckett, Supervisor, Fire Protection

T. Robinson, Fire Protection Technical Specialist

D. Scheide, Nuclear Safety and Licensing Specialist

D. Smith, Fire Protection Specialist

J. Storbakken, Reactor Operator

C. Tyrone, Manager, Quality Assurance

L. Valmonte, Consultant, Framatome

L. Young, Consultant

NRC personnel

E. Crowe, Resident Inspector, Arkansas Nuclear One

G. Mizuno, Office of General Counsel

D. Nelson, Office of Enforcement

P. Qualls, Office of Nuclear Reactor Regulation, NRC

G. Wiseman, Region II

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Open

05000313; 368/2004010-01 URI Failure to Maintain Reactor Inventory Within the Pressurizer Indicating Range (Section 1R05.4) and Inadequate Alternate Shutdown Procedure (Section 1R05.4 and 1R05.5)

-2-