IR 05000382/2007004

IR 05000382-07-004; 07/08/2007 - 10/07/2007; Waterford Steam Electric Station, Unit 3

ML073180738
Person / Time
Site:	Waterford
Issue date:	11/09/2007
From:	Clark J NRC/RGN-IV/DRP/RPB-E
To:	Walsh K Entergy Operations
References
IR-07-004
Download: ML073180738 (55)

Text

ber 09, 2007

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2007004

Dear Mr. Walsh:

On October 7, 2007, the NRC completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed report documents the inspection findings, which were discussed on October 4, 2007, with Mr. Joe Kowalewski and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents three findings of very low safety significance (Green). All of these findings were determined to involve a violation of NRC requirements. However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these violations as noncited violations (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Waterford Steam Electric Station, Unit 3, facility.

Entergy Operations, Inc. -2-In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosures, and your response, if any, will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs 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/

Jeff A. Clark, P. E.

Chief, Project Branch E Division of Reactor Projects Docket: 50-382 License: NPF-38

Enclosure:

NRC Inspection Report 050000382/2007004 w/Attachment: Supplemental Information Simplified Fire Risk Assessment for Hemyc Fire Wrap

REGION IV==

Docket No.: 50-382 License No.: NPF-38 Report No.: 05000382/2007004 Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy. 18 Killona, Louisiana Dates: July 8 through October 7, 2007 Inspectors: D. H. Overland, Acting Senior Resident Inspector G. Replogle, Senior Project Engineer G. L. Guerra, CHP, Health Physicist, Plant Support Branch G. Pick. Senior Reactor Inspector, Engineering Branch 2 P. J. Elkmann, Emergency Preparedness Inspector, Operations Branch Approved By: Jeff Clark, Chief, Project Branch E ATTACHMENTS: Supplemental Information-1- Enclosure

SUMMARY OF FINDINGS

IR05000382/2007-004; 07/08/2007 - 10/07/2007; Waterford Steam Electric Station, Unit 3;

The report covered a 3-month period of inspection by resident inspectors and a senior project engineer, a health physicist, a senior reactor inspector, and an emergency preparedness inspector. The inspectors identified three Green findings. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. 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.

Waterford 3 Plant formally committed to converting their Fire Protection Program to comply with the requirements of 10 CFR Part 50.48.(c) and National Fire Protection Association Standard 805. This involves using a risk-informed methodology. The conversion and licensing processes are expected to identify and address a variety of difficult issues that are normally the subject of triennial fire protection inspections. Since any findings in this area will be addressed under the new, rather than the existing, program, the NRC has adapted its inspection and enforcement of certain issues for plants in this situation. As a result, the scope of this inspection was modified and some issues raised in this inspection are documented but subject to enforcement discretion.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not following the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not.

The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-cutting aspect in the area of human performance, work practices component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(b)) (Section 1R19).

Green.

The inspectors identified two examples of a noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility. In the first example, the pre-fire strategy for vital switchgear Room B did not contain adequate information regarding the doors required to be open to allow the desired ventilation flowpath, nor did it contain the required number of smoke ejectors necessary to desmoke the switchgear room in a manner that would allow the implementation of OP-901-524, Fire In Areas Affecting Safe Shutdown. In the second example, the licensee did not take corrective actions for a previously identified issue in a timely fashion. Specifically, the deficiencies in the pre-fire strategy for vital switchgear Room B were first identified on August 21, 2006. The deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the non-conformance with licensee management.

The licensee entered this deficiency into their corrective action program for resolution.

The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution. Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)) (Section 4OA2).

Cornerstone: Barrier Intergrity

Green.

The inspectors identified a noncited violation of Technical Specification (TS) 3.4.7 for multiple failures to complete a radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity.

Specifically, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the TS required interval of 136 to 229 days. EBAR is the average of the sum of average beta and gamma energies per disintegration for isotopes, other than radioiodines, with half-lives greater than fifteen minutes. Daily RCS samples are compared to this calculated value in order to ensure that 10CFR50.67 dose limits at the site boundary are not exceeded in the event of an accident scenario. The licensee entered this issue into their corrective action program for resolution.

The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier. This finding had a crosscutting aspect in the area of human performance. Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)) (Section 1R22).

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status: The plant began the inspection period on July 8, 2007, at 100 percent power and remained at approximately 100 percent power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors:

(1) walked down portions of the three below listed risk important systems and reviewed plant procedures and documents to verify that critical portions of the selected systems were correctly aligned;

(2) reviewed outstanding work requests; and

(3) verified that the licensee was identifying and correcting deficiencies through their corrective action program.

• August 8, 2007: Essential chilled water system Train A

• August 27, 2007: Low-pressure safety injection system Train B

• September 5, 2007: Low-pressure safety injection system Train A Documents reviewed by the inspectors included:

• OP-009-008, Safety Injection System, Revision 19

• OP-002-004, Chilled Water System, Revision 301 The inspectors completed three samples.

b. Findings

No Findings of significance were identified.

.2 Complete Walkdown

a. Inspection Scope

The inspectors:

(1) reviewed plant procedures, drawings, the Final Safety Analysis Report, Technical Specifications, and vendor manuals to determine the correct alignment of emergency diesel generator Train A;

(2) reviewed outstanding design issues, operator work arounds, and open work requests to verify that outstanding issues

did not adversely affect the functionality of the system; and

(3) verified that the licensee was identifying and resolving equipment problems in accordance with corrective action program requirements.

Documents reviewed by the inspectors included:

• OP-009-002, Revision 301, Emergency Diesel Generator

• TM-C629.0305, Vendor Technical Manual for Cooper Bessemer Emergency Diesel Generator The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Quarterly Inspection

a. Inspection Scope

The inspectors walked down the six below listed plant areas to assess the material condition of active and passive fire protection features and their operational lineup and readiness. The inspectors:

(1) verified that transient combustibles and hot work activities were controlled in accordance with plant procedures;

(2) observed the condition of fire detection devices to verify they remained functional;

(3) observed fire suppression systems to verify they remained functional and that access to manual actuators was unobstructed;

(4) verified that fire extinguishers and hose stations were provided at their designated locations and that they were in a satisfactory condition;

(5) verified that passive fire protection features (electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory material condition;

(6) verified that adequate compensatory measures were established for degraded or inoperable fire protection features and that the compensatory measures were commensurate with the significance of the deficiency; and

(7) reviewed the Updated Final Safety Analysis Report to determine if the licensee identified and corrected fire protection problems.

• July 17, 2007: Fire Zones 8C, 11, 12, and 13

• July 19, 2007: Fire Zones RAB 15, 16, 17, 18, 19, 20, and 21

• July 23, 2007: Fire Zones RAB 33, 35, 36, 37, 38, and 39

• August 2, 2007: Fire Zones RAB 2, 23, 31, 32, and 39

• August 7, 2007: Fire Zones RAB 8B, 25, 39, Cooling Tower A, Cooling Tower B, and Fuel Handling Building

• September 12, 2007: Fire Zones RAB 1B, 8A Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

.1 Annual External Flooding

a. Inspection Scope

The inspectors:

(1) reviewed the Updated Final Safety Analysis Report, the flooding analysis, and plant procedures to assess seasonal susceptibilities involving external flooding;

(2) reviewed the Updated Final Safety Analysis Report and corrective action program to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of

(a) sump pumps,

(b) level alarm circuits,

(c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes;

(4) verified that operator actions for coping with flooding can reasonably achieve the desired outcomes; and

(5) walked down the one below listed area to verify the adequacy of:

(a) equipment seals located below the floodline,

(b) floor and wall penetration seals,

(c) watertight door seals,

(d) common drain lines and sumps,

(e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

• September 20, 2007: Susceptibility of dry cooling tower components to external flooding Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

Training Observation

a. Inspection Scope

On August 21, 2007, the inspectors observed training of senior reactor operators and reactor operators to identify deficiencies and discrepancies in the training, to assess operator performance, and to assess the evaluators critique. The training scenario

involved several instrument failures and a loss of instrument air pressure, leading to a manual reactor trip in which two control element assemblies fail to insert. An ensuing loss-of-coolant accident requires a manual initiation of safety injection and containment spray.

Documents reviewed by the inspectors included:

• Simulator Scenario Number E-83, Revision 1

• Emergency Operating Procedure OP-902-000, Revision 10, Standard Post Trip Actions

• Emergency Operating Procedure OP-902-008, Revision 14, Functional Recovery Procedure

• Emergency Operating Procedure OP-902-002, Revision 11, Loss of Coolant Accident Recovery

• Emergency Operating Procedure OP-901-511, Revision 7, Instrument Air Malfunction The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the equipment performance issue listed below to:

(1) verify the appropriate handling of structure, system, and component performance or condition problems;

(2) verify the appropriate handling of degraded structure, system, and component functional performance;

(3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of structure, system, and component issues reviewed under the requirements of the Maintenance Rule, 10 CFR Part 50 Appendix B, and the Technical Specifications.

• Safety Injection Tank leakage Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Risk Assessment and Management of Risk

a. Inspection Scope

The inspectors reviewed the four below listed assessment activities to verify:

(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and licensee procedures prior to changes in plant configuration for maintenance activities and plant operations;

(2) the accuracy, adequacy, and completeness of the information considered in the risk assessment;

(3) that the licensee recognizes, and/or enters as applicable, the appropriate licensee-established risk category according to the risk assessment results and licensee procedures;

(4) the licensee properly controlled emergent work; and

(5) the licensee identified and corrected problems related to maintenance risk assessments.

• August 22, 2007: Planned maintenance outage of shield building ventilation Train A

• August 27, 2007: Planned maintenance outage of low-pressure safety injection Train A

• September 11, 2007: Planned maintenance outage of emergency feedwater Train B

• September 14, 2007: Planned surveillance activities for undervoltage and shunt trip coil testing for reactor trip circuit breakers Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors:

(1) reviewed plants status documents such as operator shift logs, emergent work documentation, deferred modifications, and standing orders to determine if an operability evaluation was warranted for degraded components;

(2) referred to the Updated Final Safety Analysis Report and design-basis documents to review the technical adequacy of licensee operability evaluations;

(3) evaluated compensatory measures associated with operability evaluations;

(4) determined degraded component impact on any Technical Specifications;

(5) used the Significance

Determination Process to evaluate the risk significance of degraded or inoperable equipment; and

(6) verified that the licensee has identified and implemented appropriate corrective actions associated with degraded components.

• July 20, 2007: Operability evaluation addressing Appendix R required emergency lighting

• July 25, 2007: Operability evaluation addressing a crack in the Fuel Handling Building ceiling

• July 27, 2007: Operability evaluation addressing pressurizer heater design

• August 1, 2007: Operability evaluation addressing dry cooling tower Train A sump pump low flow

• August 2, 2007: Operability evaluation addressing recurring primary side steam generator valve and loose parts monitor system alarms Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed five samples.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors selected the six below listed postmaintenance test activities of risk significant systems or components. For each item, the inspectors:

(1) reviewed the applicable licensing basis and/or design-basis documents to determine the safety functions;

(2) evaluated the safety functions that may have been affected by the maintenance activity; and

(3) reviewed the test procedure to ensure it adequately tested the safety function that may have been affected. The inspectors either witnessed or reviewed test data to verify that acceptance criteria were met, plant impacts were evaluated, test equipment was calibrated, procedures were followed, jumpers were properly controlled, the test data results were complete and accurate, the test equipment was removed, the system was properly realigned, and deficiencies during testing were documented. The inspectors also reviewed the Updated Final Safety Analysis Report to determine if the licensee-identified and corrected problems related to postmaintenance testing.

• August 9, 2007: Corrective maintenance to replace Transducers 5 through 8 on main feedwater ultrasonic flow meter Number 2

• August 29, 2007: Corrective maintenance to replace a faulty relay in essential chiller Train A chilled water Pump 1 breaker

• September 18, 2007: Planned maintenance to stroke test containment atmospheric purge Valves 103 and 104 following breaker maintenance

• September 19, 2007: Planned maintenance to stroke test emergency feedwater Valve 228A following breaker maintenance

• September 20, 2007: Planned maintenance on emergency feedwater Pump A to change the oil and lube the pump

• September 12, 2007: Planned maintenance to clean boric acid from containment spray Pump B Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

Introduction.

The inspectors identified a Green noncited violation of Technical Specification 6.8.1.a (Procedures) for an inadequate boric acid evaluation procedure and for the failure to follow the same procedure. Specifically, the procedure noted that small amounts of boric acid could severely corrode carbon and low alloy carbon steel, but only had engineers check drawings for carbon steel components. Components with low alloy steel on the containment spray pumps were sometimes ignored. In addition, the procedure required pictures of the boric acid condition but, for some evaluations, no pictures were taken of the containment spray pump leaks. This made trending of the condition, to check for worsening, difficult. The inspectors determined that engineers were not reading the boric acid evaluation procedure when performing the evaluations, they simply filled out the forms. The procedure contained valuable insights vital for proper boric acid evaluations, whereas the forms did not.

Description.

During a plant tour on September 12, 2007, the inspectors observed boric acid deposits in the Train A and B containment spray pump shaft cavities. The deposits originated from the mechanical shaft seal area and collected on various components, including low alloy steel bolts on the mechanical seal assembly. The inspectors asked the licensee for the boric acid evaluations which were required by Procedure EN-DC-319, "Inspection and Evaluation of Boric Acid Leaks," Revision 0. The latest evaluation for Pump B was dated May 25, 2007 and the most recent evaluation for Pump A was dated March 10, 2006.

NOTE: The licensee currently performs boric acid evaluations in accordance with Revision 1 of Procedure EN-DC-319. In addition, this report discusses historical boric acid evaluations performed in accordance with other versions of the procedure. With respect to the procedural content related to this violation, there were no meaningful differences between the procedures.

Inadequate Procedure: The inspectors identified that Procedure EN-DC-319 was inadequate, in that it did not require actions to address boric acid wastage on low alloy

steels. The forms used for the evaluation, "Attachment 9.3, Identification of Boric Acid Leakage," and "Attachment 9.4, Evaluation/Screening of Boric Acid Leakage," did not require the engineers to identify or evaluate the impact of boric acid on low alloy steels.

The forms only required actions to address carbon steel components. Engineers that performed the evaluations stipulated that there was a difference between carbon steel and low alloy steel and they were not required by the procedure (forms) to evaluate the latter.

Low alloy steel is carbon steel with low amounts of selected alloys added to enhance material hardening characteristics. The small amount of alloys do not enhance resistance to boric acid wastage. For example, NRC Information Notice 80-27, "Degradation of Reactor Coolant Pump Studs," details an instance where another utility identified significant boric acid wastage of reactor coolant pump low allow steel closure studs.

Further, Procedure EN-DC-319 contained the following precaution:

Small amounts of boric acid have the potential to severely corrode high temperature carbon and low allow steel over a long period of time.

The inspectors had noted that boric acid was in contact with low alloy steel bolts on both containment spray pump mechanical seal housings but the boric acid evaluations did not address this condition. The pattern of dry boric acid suggested that the leaks were traversing past and onto the bolt shanks and threads, which were not visible unless removed from the assembly.

The licensee did have some pictures for containment spray Pump B (but not for A) and the inspectors noted that the current Pump B boric acid pattern was consistent with pictures dated May 25, 2007. Since engineers did not consistently take pictures and documented leak descriptions were lacking detail, the impact and duration of the leaks was difficult to determine. At the present, the inspectors noticed some, but very limited, evidence of material wastage. Some bolt heads showed small amounts of external corrosion, but dried leakage past the bolt internals did not appear discolored.

Therefore, currently pump operability was not in question.

The broader concern was that the same boric acid evaluation forms were used for all boric acid leaks and some components were more vulnerable to faulty evaluations than others. For example, components in containment are not readily inspectible and hot boric acid leaks on low alloy steel components could result in much more significant, but unaddressed, wastage. Therefore, the inspectors determined that this violation would be more significant if left uncorrected.

Pictures: The inspectors identified that the boric acid evaluator for the Train A, March 10, 2006 evaluation had failed to follow Procedure UNT-006-031, Revision 0 (a previous version of EN-DC-319), in that no pictures were taken to describe the condition.

Procedure UNT-006-031 specified, in part:

If possible, include pictures of the overall component, close-up leakage conditions or any other relevant condition to assist in describing the condition and component location.

Contrary to the above, it was possible to include pictures of the overall component and boric acid build up but no pictures were taken. In addition, the description on the licensee's boric acid evaluation form was vague and it was impossible to tell if the leak had gotten worse or if the buildup of boric acid was stagnant for a sustained period.

Overall Assessment: While the licensee's procedure provided valuable information regarding the vulnerability of low alloy steel to boric acid wastage and the need to document boric acid leaks with pictures, engineers did not routinely follow the procedure. Instead, they accomplished their evaluations by simply filling out the form.

The failure to properly use and implement the procedure was a significant contributor to the violation and its significance (more than minor).

Analysis.

The failure to establish an adequate procedure for boric acid evaluations, and the failure to implement the procedure, were performance deficiencies. The finding was more than minor because it could, if left uncorrected, result in a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green)because it did not result in an actual loss of safety function for the containment spray system. The cause of the finding has a cross-cutting aspect in the area of human performance, resources component, in that the licensee failed to effectively communicate the expectations regarding procedural compliance and personnel follow procedures (H.4(c)).

Enforcement.

Technical Specification 6.8.1.a (Procedures) requires the licensee to establish and implement procedures recommended by Appendix A to Regulatory Guide 1.33, Revision 2, 1978. Appendix A, Section 9 recommends procedures for maintenance, including inspection. Procedure UNt-006-031 required, in part, pictures of boric acid leaks, if possible. Contrary to the above, it was possible to take pictures of the boric acid leak evaluated by the March 10, 2006 but no pictures were taken. In addition, Procedure EN-DC-319 was inadequate, in that it did not require engineers to evaluate the impact of boric acid on low alloy steels. Because this finding was of very low safety significance and has been entered into the licensees corrective action program (CR-WF3-2007-03590), it is considered a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000382/2007004-01, Inadequate Boric Acid Leak Evaluations.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, procedure requirements, and Technical Specifications to ensure that the five below listed surveillance activities demonstrated that the structures, systems, and components tested were capable of performing their intended safety functions. The inspectors either

witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate:

(1) preconditioning;

(2) evaluation of testing impact on the plant;

(3) acceptance criteria;

(4) test equipment;

(5) procedures;

(6) jumper/lifted lead controls;

(7) test data;

(8) testing frequency and method demonstrated Technical Specification operability;

(9) test equipment removal;

(10) restoration of plant systems;

(11) fulfillment of ASME Code requirements;

(12) updating of performance indicator data;

(13) engineering evaluations, root causes, and bases for returning tested structures, systems, and components not meeting the test acceptance criteria were correct;

(14) reference setting data; and

(15) annunciators and alarms setpoints. The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing.

• July 26, 2007: Maintenance Procedure MM-007-010, Revision 15, Change 3, Fire Extinguisher Inspection and Replacement, is used to ensure that all site fire extinguishers are in working condition.

• August 7, 2007: Surveillance Procedure OP-903-003, Revision 11, Change 1, Charging Pump Operability Check, is used to ensure that charging Pump AB discharge pressure, flow, and vibration characteristics are within design parameters.

• August 7, 2007: Surveillance Procedure OP-903-035, Revision 12, Containment Spray Pump Operability Check, is used to ensure that containment spray Pump B discharge pressure, flow, and vibration characteristics are within design parameters.

• August 20, 2007: Surveillance Procedure OP-903-046, Revision 301, Emergency Feedwater Pump Operability Check, is used to ensure that emergency feedwater Pump AB discharge pressure, flow, and vibration characteristics are within design parameters.

• September 5, 2007: Chemistry Procedure CE-003-306, Revision 9, Determination of the Average Beta-Gamma Energy of Reactor Coolant, calculates the activity in the reactor coolant due to radioisotopes with a half-life of greater than 15 minutes.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed five samples.

b. Findings

Introduction.

The inspectors identified a Green noncited violation of Technical Specification Surveillance Requirement 4.4.7 for multiple failures to complete a radiochemical analysis for EBAR (Average Disintegration Energy) determination within the required periodicity.

Description.

Technical Specification Surveillance Requirement 4.4.7 requires:

The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

Table 4.4-4, Primary Coolant Specific Activity Sample and Analysis Program item 3, requires a radiochemical analysis for EBAR determination to be completed once every 6 months.

EBAR is the average (weighted in proportion to concentration of each radionuclide in reactor coolant at time of sampling) of the sum of average beta and gamma energies per disintegration (in MeV) for isotopes, other than radioiodines, with half-lives greater than 15 minutes, making up at least 95 percent of total noniodine activity in reactor coolant. The EBAR value is then divided into a correction factor and utilized in a calculation to generate a value that establishes a maximum reactor coolant system (RCS) activity limit in microcuries per milliliter. Daily RCS samples are compared to this calculated value in order to ensure that 10 CFR 50.67 dose limits at the site boundary are not exceeded during an accident scenario.

On February 26, 2007, the RCS was sampled for EBAR relevant isotopes. Per procedure, several strontium and iron isotope samples were sent offsite for analysis.

Results were complied and on May 1, 2007, the EBAR calculation was performed as a training performance evaluation for qualification of a chemistry technician. On September 5, 2007, the licensee noticed that the EBAR calculation had never been reviewed and the calculated value was never implemented for daily comparison.

Condition Report CR-WF3-2007-3146 was generated.

In response to further questioning by the senior resident inspector about extent of condition, the licensee discovered that although the EBAR reactor coolant samples dating back to December 1999 were drawn on time, on thirteen out of fifteen occasions, the licensee had failed to complete the analysis and replace the old EBAR value with the new EBAR value within the Technical Specification-required interval of 136 to 229 days.

On the two occasions that the Technical Specification requirement was met, it was only met due to the allowance of the 25 percent grace period. The average time for an EBAR value to be in place was 284 days, with the longest time period lasting 566 days.

The ability to sample, but fail to complete the analysis on time was due, in part, to the tracking method in place. A task to collect the EBAR sample is generated during the required periodicity. However, once the sample is obtained, the task is marked as complete and there are no additional tasks to ensure that the analysis of the sample or results calculation are completed.

Analysis.

The failure to follow plant technical specifications and properly sample and analyze reactor coolant system chemistry to calculate a current EBAR value was a performance deficiency. The finding was determined to be NRC identified because it involved a previously documented licensee finding to which the inspector significantly added value. The finding was more than minor because it was associated with the cladding performance attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers

(fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance (Green) because it only affected the fuel barrier.

This finding had a crosscutting aspect in the area of human performance. Specifically, the licensees personnel work practices failed to support human performance by ensuring that activity status and completion are properly documented (H.4(a)).

Enforcement:

Surveillance Requirement 4.4.7 of Technical Specification 3.4.7 requires a radiochemical analysis for EBAR determination to be completed once every 6 months.

Contrary to the above, on thirteen different occasions between January 2000 and September 2007, radiochemical analyses for EBAR determination were not properly conducted. Because this finding was of very low safety significance and has been entered into the licensees corrective action program (CR-WF3-2007-3301), it is considered a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000382/2007004-02, Missed Reactor Coolant System Chemistry Samples.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspector performed an in-office review of Revision 22 to the Waterford 3 Emergency Plan Implementing Procedure EP-001-001, "Recognition and Classification of Emergency Conditions," Revision 22, received September 7, 2007. This revision added information about the choice of meteorological instruments used to measure the wind speed to the basis for Emergency Action Level HA6.

The revision was compared to its previous revision, to the criteria of NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, to Nuclear Energy Institute report 99-01, Methodology for Development of Emergency Action Levels, Revision 4, to the requirements of 10 CFR 50.47(b), and to 50.54(q) to determine if the licensee adequately implemented 10 CFR 50.54(q). This review was not documented in a safety evaluation report and did not constitute approval of licensee changes, therefore the changes are subject to future inspection.

The inspector completed one sample during this inspection.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance.

During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation packages reported by the licensee in the Occupational Radiation Safety Cornerstone

• Controls (surveys, posting, and barricades) of radiation, high radiation, and airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler locations

• Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms.

• Barrier integrity and performance of engineering controls in airborne radioactivity areas

• Adequacy of the licensees internal dose assessment for any actual internal exposure greater than 50 millirem Committed Effective Dose Equivalent

• Physical and programmatic controls for highly activated or contaminated materials (nonfuel) stored within spent fuel and other storage pools.

• Self-assessments, audits, licensee event reports, and special reports related to the access control program since the last inspection

• Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individual deficiencies

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls such as, required surveys, radiation protection job coverage, and contamination controls during job performance

• Dosimetry placement in high radiation work areas with significant dose rate gradients

• Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas

• Controls for special areas that have the potential to become very high radiation areas during certain plant operations

• Posting and locking of entrances to all accessible high dose rate - high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements The inspector completed 21 of the required 21 samples.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspector used the requirements in 10 CFR Part 20 and the licensees procedures required by technical specifications as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:

• Current 3-year rolling average collective exposure

• Work activities of exposure significance completed during the last outage

• ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements

• Integration of ALARA requirements into work procedure and radiation work permit documents

• Shielding requests and dose/benefit analyses

• Dose rate reduction activities in work planning

• Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding

• Source-term control strategy or justifications for not pursuing such exposure reduction initiatives

• Specific sources identified by the licensee for exposure reduction actions and priorities established for these actions, and results achieved against since the last refueling cycle

• Declared pregnant workers during the current assessment period, monitoring controls, and the exposure results The inspector completed 10 of the required 29 samples.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

Cornerstone: Mitigating Systems

The inspectors sampled licensee submittals for the three mitigating system performance index indicators listed below for the period of January 2006 through September 2007.

The definitions and guidance of Nuclear Energy Institute 99-02, Regulatory Assessment Indicator Guideline, Revision 4, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of performance indicator data reported during the assessment period. The inspectors reviewed licensee event reports, out-of-service logs, operating logs, and the Maintenance Rule database as part of the assessment. Licensee performance indicator data were also reviewed against the requirements of Procedure EN-LI-114, Performance Indicator Process, Revision 2.

• Emergency AC Power

• Support Cooling Water Systems

• Safety System Functional Failures Occupational Radiation Safety Cornerstone The inspector reviewed licensee documents from October 1, 2006, through June 30, 2007. The review included corrective action documentation that identified occurrences in locked high radiation areas (as defined in the licensees technical specifications), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in Nuclear Energy Institute (NEI) 99-02). Additional records reviewed included as low as reasonably achievable (ALARA) records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator data. In addition, the inspector toured plant areas to verify that high radiation, locked high radiation, and very high radiation areas were properly controlled. Performance indicator definitions and guidance contained in NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 4, were used to verify the basis in reporting for each data element.

• Occupational Exposure Control Effectiveness

The inspector completed the required sample

(1) in this cornerstone.

Public Radiation Safety Cornerstone The inspector reviewed licensee documents from October 1, 2006, through June 30, 2007. Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded performance indicator thresholds and those reported to the NRC. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator data. Performance indicator definitions and guidance contained in NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 4, were used to verify the basis in reporting for each data element.

• Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into the licensees corrective action program. This assessment was accomplished by reviewing condition reports and event trend reports and attending daily operational meetings. The inspectors:

(1) verified that equipment, human performance, and program issues were being identified by the licensee at an appropriate threshold and that the issues were entered into the corrective action program;

(2) verified that corrective actions were commensurate with the significance of the issue; and

(3) identified conditions that might warrant additional followup through other baseline inspection procedures.

b. Findings

No findings of significance were identified.

.2 Selected Issue Followup Inspection

a. Inspection Scope

In addition to the routine review, the inspectors selected the three issues, listed below, for a more in-depth review. The inspectors considered the following during the review of the licensees actions:

(1) complete and accurate identification of the problem in a timely manner;

(2) evaluation and disposition of operability/reportability issues;

(3) consideration of extent of condition, generic implications, common cause, and previous occurrences;

(4) classification and prioritization of the resolution of the problem;

(5) identification of root and contributing causes of the problem;

(6) identification of corrective actions; and

(7) completion of corrective actions in a timely manner.

• September 18, 2007: Feasibility of manual compensatory actions in vital switchgear Room AB during a fire in vital switchgear Room B

• Access Control to Radiologically Significant Areas (Section 2OS1)

• ALARA Planning and Controls (Section 2OS2)

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed three samples.

b. Findings

Introduction.

The inspectors identified two examples of a Green noncited violation of Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 for failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility.

Description.

Updated Final Safety Analysis Report (UFSAR) Section 9.5.1.3.1 provides a comparison of the licensees Fire Protection Program to criteria described in Branch Technical Position APCSB 9.5-1, Revision 0, Appendix A.

.1 UFSAR Section 9.5.1.3.1.B.1, Administrative Procedures, Controls and Fire

Brigade, describes Branch Technical Position APCSB 9.5-1, Revision 0, Appendix A requirement that:

Administrative procedures consistent with the need for maintaining the performance of the fire protection system and personnel in nuclear power plants should be provided.

The licensees response to the requirement listed above refers to licensee Procedure UNT-050-013, Fire Protection Program. Procedure UNT-050-013 states that prefire strategy format and content requirements are described in licensee Procedure FP-001-018, Pre-Fire Strategies, Development and Revision. Procedure FP-001-018, Section 6.1.3 states that:

Pre-fire strategies should include ... ventilation system operation that ensures desired plant air distribution when the ventilation flow is modified for fire containment or smoke clearing operations.

Contrary to the above requirement, the prefire strategy for vital switchgear Room B (fire zone RAB 8B) did not contain adequate information regarding the doors required to be open to allow the desired ventilation flowpath (Door 11), nor did it contain the required number of smoke ejectors

(2) necessary to desmoke the switchgear room in a manner that would allow the implementation of Procedure OP-901-524, Fire In Areas Affecting Safe Shutdown. Specifically, a manual action, which serves as a compensatory measure for the licensees

noncompliance with UFSAR Section 9.5.1.3.1.D.1.(a) requirements for separation of safe shutdown trains would not be feasible based on smoke levels in vital switchgear Room AB (fire zone RAB 8C). Switchgear Room AB is located immediately next to switchgear Room B, and a fire in one of the rooms would allow smoke to enter the other room due to a large opening at the top of the wall that separates the switchgear rooms. For the manual action in switchgear Room AB to be feasible, switchgear Room B would need to be desmoked within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Per Calculation ECF-05-003, Manual Action Fire Model - RAB 8, the only way to ensure that switchgear Room AB is habitable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is to desmoke switchgear Room B with two smoke ejectors and by ensuring that Door 11 remains open to allow a supply of fresh air to replace the smoke being ejected out to Dry Cooling Tower B area through Door 51.

.2 UFSAR Section 9.5.1.3.1.C.8, Corrective Action describes Branch Technical

Position APCSB 9.5-1, Revision 0, Appendix A requirement that:

Measures should be established to assure that conditions adverse to fire protection, such as failures, malfunctions, deficiencies, deviations, defective components, uncontrolled combustible material and nonconformances are promptly identified, reported, and corrected.

The licensees response to the requirement listed above refers to the Quality Assurance Program Manual (Special Scope) and the Fire Protection Program.

Procedure UNT-050-013, Fire Protection Program, contains no discussion of corrective action criteria. The Quality Assurance Program Manual (Special Scope) Section 5.10, Corrective Action, refers to Site Directive W2.501, Corrective Action. Site Directive W2.501, and the corrective action criteria contained therein, was subsumed by Procedure EN-LI-102, Corrective Action Process. Procedure EN-LI-102, Section 5.8 [2](f) requires, in part, that corrective actions be timely. Procedure EN-LI-102, Attachment 9.4, Corrective Action Processing Guidelines, directs that a Category C condition report corrective action should be, corrected within a timeframe specified by the CRG (normally less than 180 days).

The noncompliance discussed with UFSAR Section 9.5.1.3.1.B.1 was first identified as a potential vulnerability in Condition Report CR-2006-2407, dated August 21, 2006, and a corrective action recommendation was made to revise the deficient procedure. On August 28, 2006, Condition Report CR-2006-2407 was closed to Condition Report CR-2006-0388 corrective action CA-4, which took action to consider the need to revise the pre-fire plan. On September 5, 2006, Condition Report CR-2006-0388 corrective action CA-5 was created to revise the pre-fire plan. On February 12, 2007, Condition Report CR-2006-0388 corrective action CA-5 was closed to Condition Report CR-2007-0346 corrective action CA-17, which had a due date of December 31, 2008. Contrary to the above corrective action timeliness requirement, the deficient procedure was not corrected until September 14, 2007, after the senior resident inspector discussed the nonconformance with licensee management.

Analysis.

The licensees failure to follow Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 and implement their fire protection program as

described in the UFSAR is a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Protection Against External Factors) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix F, Phase 1 initial qualitative screening, the issue screened as having very low safety significance because the compensatory manual action required to safely shut down the plant is not needed in order to reach hot shutdown. This finding had a crosscutting aspect in the area of problem identification and resolution.

Specifically, the licensees personnel corrective action process failed to take appropriate corrective actions to address the safety issue in a timely manner (P.1(d)).

Enforcement:

Waterford Steam Electric Station, Unit 3 Facility Operating License Condition 2.C.9 requires that the licensee implement their fire protection program as described in the UFSAR. Contrary to the above, the licensee failed to maintain an adequate prefire strategy procedure as described in the UFSAR. Also, contrary to the above, the licensee failed to follow the corrective action program as described in the UFSAR. Because this finding was of very low safety significance and has been entered into the licensees corrective action program (Condition Report CR-WF3-2007-3264), it is considered a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000382/2007004-03, Inadequate Procedure for a Fire in Vital Switchgear Room B.

.3 Semiannual Trend Review

a. Inspection Scope

The inspectors completed a semiannual trend review of repetitive or closely related issues associated with the Appendix R required emergency lights to identify trends that might indicate the existence of more safety significant issues. The inspectors review consisted of the 4 year period between January 2003 to September 2007. When warranted, some of the samples expanded beyond those dates to fully assess the issue.

The inspectors also reviewed corrective action program items associated with troubleshooting. The inspectors compared and contrasted their results with the results contained in the licensees quarterly trend reports. Corrective actions associated with a sample of the issues identified in the licensees trend report were reviewed for adequacy.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

(Closed) Unresolved Item 05000382/2000007-02: Determine the Qualification of Heymc Fire Wrap as a 1-hour Rated Fire Barrier

a. Inspection Scope

The team had opened this item because of questions related to the acceptability of the tested configuration (4-inch conduits) versus the installed configurations (1- and 2-inch conduits) in Fire Area RAB-2 and indicated further NRC review was required.

An NRC inspector and a senior reactor analyst performed an in-office review of the licensee's interim measures and risk assessment to determine if the licensee had demonstrated that the significance of the issue was less than high safety significance (Red). The inspector performed this inspection by reviewing the documents listed in the attachment and discussed below. The inspector and senior reactor analyst discussed the issues with the fire protection engineer and licensee probabilistic safety assessment personnel.

The inspector performed the evaluation in this manner because Waterford 3 formally committed to converting their Fire Protection Program to comply with the requirements of 10 CFR 50.48(c) and NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, prior to December 31, 2005. This involves using a risk-informed methodology. The conversion and licensing processes are expected to identify and address a variety of difficult issues that are normally the subject of triennial fire protection inspections. Since any findings in this area will be addressed under the new, rather than the existing, program, the NRC has adapted its inspection and enforcement of certain issues for plants in this situation.

b. Findings

Introduction.

The inspector identified an apparent violation of License Condition 2.C.9 because the licensee failed to maintain adequate separation between redundant trains of safe shutdown equipment. Specifically, NRC had determined that the installed Heymc fire barrier material can not provide the required 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of protection. However, this violation will not be cited since the licensee met the Enforcement Policy criteria for enforcement discretion for a plant committed to adopting NFPA Standard 805.

Description.

When identified in Calendar Year 2000, the team determined that the installed conduits did not match the tested configuration described in the fire test report for the Heymc fire wrap material. The licensee used the Heymc fire wrap as a 1-hour fire rated barrier to separate safe shutdown functions within the same fire area. The laboratory had performed testing of 4-inch diameter Heymc-wrapped conduits; however, the team identified 1- and 2-inch diameter conduits containing safe-shutdown cables wrapped with Heymc.

The NRC conducted testing of Heymc material and documented the test results in Information Notice 2005-07, "Results of Heymc Electrical Raceway Fire Barrier System Full Scale Fire Testing." NRC tested the following four common methods of joining the Heymc material into a complete electrical raceway fire barrier system:

(1) using stitched joints,

(2) using minimum 6-inch collars over a joint,

(3) using minimum 2-inch overlapping of the mats, and

(4) using through bolts with fender washers. The information notice describes the impact upon each of the methods, which resulted in opening of each of the joint systems and exposing the assembly (conduit, cable tray,

junction box and air drop cable) to the furnace environment. The testing demonstrated that all but one assembly (conduit or cable tray) experienced temperatures capable of damaging plant cables as identified in Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 7.

In response to Information Notice 2005-007, the licensee initiated compensatory measures and initiated Condition Report 2005-01178. As immediate corrective actions, the licensee initiated hourly fire tours of the 19 fire areas that contained the Heymc material. The licensee determined in their apparent cause evaluation that the presence of sprinklers (except for two areas with approved deviations from Appendix R) and the presence of their fire brigade ensures that a fire would not impact the plants ability to perform a safe shutdown following a fire. As long-term corrective actions, the licensee developed a Heymc Resolution Action Plan that included: identifying the material locations and configurations, testing the configurations, initiating plans and training for replacing the Heymc as needed with an approved fire barrier material, and performing a study to identify options for addressing the Heymc since replacing all of the Heymc was identified as cost prohibitive.

NRC issued Generic Letter 2006-03, "Potentially Nonconforming Heymc and MT Fire Barrier Configurations," to require that licensees evaluate their facilities to confirm compliance with existing regulations. Specifically, Generic Letter 2006-03 required licensees to discuss the installation of Heymc or MT barrier materials and the impact on their facility including whether the installation was described in their licensing basis. The generic letter further required a description of their corrective actions and planned completion date.

The licensee described in their Generic Letter 2006-03 response that they had Heymc installed extensively throughout the facility on conduits, cable trays, containment penetrations, and inside containment as a radiant energy shield. Because of the estimated cost to replace the Heymc, the licensee elected to adopt NFPA 805, in accordance with 10 CFR 50.48(c). The licensee described their intent to adopt NFPA 805 by letter dated December 21, 2005. Because of the time required to transfer to an NFPA 805 based fire protection program, the licensee indicated they would not have all corrective actions completed by December 2008.

Analysis.

Failure to meet the separation requirements for a 1-hour fire barrier was a performance deficiency since the licensee did not comply with their Fire Protection Program, as required by License Condition 2.C.9. This finding was more than minor because it affected the protection against external factors attribute of the Mitigating Systems cornerstone. As specified in the enforcement policy, the licensee had performed a simplified risk assessment, included as Attachment B. The inspectors reviewed the simplified fire area-by-fire area risk assessment and determined that the licensee demonstrated that the risk was less than high safety significance (Red).

Enforcement.

License Condition 2.C.9, states, in part, that the licensee shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility through Amendment 36 and as approved in the Safety Evaluation Report through Supplement 9. Section 9.5.1.4(1)of the Waterford Safety Evaluation Report states that the licensee committed to provide 1-hour fire rated barriers to protect one division of shutdown-related cables in cable trays

and conduits in certain fire areas. Contrary to the above the licensee had installed an inadequate 1-hour fire barrier in 19 different fire areas that could have impacted the ability to safely shutdown the facility. The licensee included this item in their corrective action program as Condition Report 2005-01178.

Because the licensee committed to adopting NFPA Standard 805 and changing their fire protection program license basis to comply with 10 CFR 50.48.(c), this issue is covered by enforcement discretion in accordance with the NRC Enforcement Policy. Specifically, the licensee:

(1) would have identified and addressed this issue during the conversion to NFPA Standard 805,

(2) had entered this issue into their corrective action program and implemented appropriate compensatory measures,

(3) demonstrated the finding would not be categorized under the Reactor Oversight Process as Red or a Severity Level I violation, and

(4) submitted their letter of intent prior to December 31, 2005. The inspector determined that this violation meets the criteria for enforcement discretion for plants in transition to a risk-informed, performance-based fire protection program as allowed per 10 CFR Part 50.48(c). Since all the criteria were met, the NRC is exercising enforcement discretion for this issue.

4OA6 Meetings, Including Exit

.1 On August 16, 2007, the inspector presented the occupational radiation safety inspection

results to Mr. J. Kowalewski and other members of your staff who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

.2 On September 20, 2007, the inspectors discussed the results of their review with

Mr. O. Pipkins, Senior Licensing Engineer. The inspectors returned all proprietary information to the licensee.

.3 On September 19, 2007, the emergency preparedness inspector conducted a telephonic

exit meeting to present the inspection results to Mr. J. Lewis, Manager, Emergency Preparedness, who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

.4 On October 4, 2007, the resident inspectors presented the inspection results to

Mr. Joe Kowalewski and other members of licensee management at the conclusion of the inspection. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Anders, Superintendent, Plant Security

H. Brodt, Risk Analyst

K. Cook, Director, Nuclear Safety Assurance

L. Dauzat, Supervisor, Radiation Protection

A. Dodds, Manager, Operations

G. Fey, Planning and Scheduling

J. Kowalewski, General Manager, Entergy

J. Lewis, Manager, Emergency Preparedness

D. Marpe, Project Manager

M. Mason, Technical Specialist, Licensing

C. Miller, Assistant Manager, Radiation Protection

R. Murillo, Manager, Licensing

D. Newman, Supervisor, Radiation Protection

K. Nichols, Director, Engineering

B. Pilutti, Manager, Radiation Protection

R. Putnam, Manager, Programs and Components

S. Ramzy, Engineer, Radiation Protection

G. Scott, Engineer, Licensing

K. T. Walsh, General Manager, Plant Operations

Attachment

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000382/2007004-01 NCV Inadequate Boric Acid Leak Evaluations(Section 1R19)

05000382/2007004-02 NCV Missed Reactor Coolant System Chemistry Samples (Section 1R22)

05000382/2007004-03 NCV Inadequate Procedure for a Fire in Vital Switchgear Room B (Section 4OA2)

05000382/2007004-04 AV Determination as to the qualification of Heymc fire wrap as a rated 1-hour fire barrier (Section 4OA5)

Closed

05000382/2007004-01 NCV Inadequate Boric Acid Leak Evaluations (Section 1R19)

05000382/2007004-02 NCV Missed Reactor Coolant System Chemistry Samples (Section 1R22)

05000382/2007004-03 NCV Inadequate Procedure for a Fire in Vital Switchgear Room B (Section 4OA2)

05000382/2000007-02 URI Determination as to the qualification of Heymc fire wrap as a rated 1-hour fire barrier (Section 4OA5)

05000382/2007004-04 AV Determination as to the qualification of Heymc fire wrap as a rated 1-hour fire barrier (Section 4OA5)

Attachment