Text

IR 05000454/2006004; 05000455/2006004; on 07/01/2006 - 09/30/2006; Byron Station, Units 1 and 2; Fire Protection and Radiation Protection
	ML063120614
Person / Time
Site:	Byron
Issue date:	11/08/2006
From:	Richard Skokowski; NRC/RGN-III/DRP/RPB3
To:	Crane C; Exelon Generation Co, Exelon Nuclear
References
	FOIA/PA-2010-0209 IR-06-004
	Download: ML063120614 (60)
	v • d • e

November 8, 2006 Mr. Christopher President and Chief Nuclear Officer

Exelon Nuclear

Exelon Generation Company, LLC

4300 Winfield Road

Warrenville, IL 60555SUBJECT:BYRON STATION, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000454/2006004; 05000455/2006004

Dear Mr. Crane:

On September 30, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Byron Station, Units 1 and 2. The enclosed report documents the

inspection findings which were discussed on October 03, 2006, with Mr. Dave Hoots and other

members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's 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 NRC-identified findings of very low safety significance (Green). All three findings involved violations of NRC requirements. In addition, two licensee-identified

violations which were determined to be of very low safety significance are listed in this report.

However, because of the very low safety signi ficance of the violations and because they were entered into your corrective action program, the NRC is treating these violations as non-cited

violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, 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, D.C.

20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -

Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of

Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the

Resident Inspector office at the Byron facility.

C. Crane-2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC

Public Document Room or from the Pub licly 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 PLouden acting for/

Richard A. Skokowski, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66Enclosure:Inspection Report 05000454/2006004; 05000455/2006004; w/Attachments 1: Supplemental Information 2: Confirmatory Measurements Comparison Chart

3: Tritium Sample Resultscc w/encl:Site Vice President - Byron Station Plant Manager - Byron Station

Regulatory Assurance Manager - Byron Station

Chief Operating Officer

Senior Vice President - Nuclear Services

Vice President - Mid-West Operations Support

Vice President - Licensing and Regulatory Affairs

Director Licensing

Manager Licensing - Braidwood and Byron

Senior Counsel, Nuclear

Document Control Desk - Licensing

Assistant Attorney General

Illinois Emergency Management Agency

State Liaison Officer, State of Illinois

State Liaison Officer, State of Wisconsin

Chairman, Illinois Commerce Commission

B. Quigley, Byron Station

SUMMARY OF FINDINGS

IR 05000454/2006004; 05000455/2006004; on 07/01/2006-09/30/2006; Byron Station, Units 1 and 2; Fire Protection; and Radiation Protection.

This report covers a 3-month period of baseline resident inspection and announced baseline inspections on Radiation Protection and Temporary Instruction (TI) 2515/166, "Pressurized

Water Reactor Containment Sump Blockage." These inspections were conducted by four regional inspectors and the resident inspectors. Three Green findings, all of which were non-

cited violations (NCV), were identified. 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. 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.A.Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Non-Cited Violation of Byron Facility Operating License Nos. NPF-37 and NPF-66, Condition 2.c.6, for failing to maintain the firewall separating the Auxiliary Building from the penetration area in accordance with the approved fire protection program. Fire seals were required to be provided in this firewall, except where an evaluation had been perfo rmed and approved to allow a deviation.

Two sleeves containing fire seals had pulling ropes embedded in the fire seals in the firewall separating the Auxiliary Building General Area 401 from the Unit 1 piping penetration area; also, no evaluation or exemption existed to justify this configuration.

The licensee entered the issue into its corrective action program for resolution and implemented compensatory measures that included hourly fire watches.

This finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure that external factors (i.e., fire, flood, etc) do not impact the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because the fire seals were in small diameter sleeves that traveled a distance of 45 feet and had two 90 degree bends and the location of combustibles were positioned such that the piping penetration end of the fire seals would not be subject to direct flame impingement. (Section 1R05)

Cornerstone: Occupational Radiation Safety

Green.

An inspector-identified finding of very low safety significance and two associated Non-Cited Violations of NRC requirements were identified for the failure to post and control access to High Radiation Areas, as required by 10 CFR Part 20, to notify individuals of the radiological hazard present and to prevent the unauthorized entry to such areas. Specifically, the entrance to the Unit 1 Filter Valve Aisle located on the 383'

elevation of the Auxiliary Building, a high radiation area with a radiation dose rate of approximately 135 millirem in one hour, was not posted or controlled by any of the methods described in 10 CFR 20.1902, 10 CFR 20.1601, or Technical Specification 5.7.1.

The issue was more than minor because the issue was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The issue represents a finding of very low safety significance because the finding did not constitute an ALARA or work control issue, did not result in an overexposure or the substantial potential for an overexposure, and did not compromise the licensee's ability to assess dose. Non-Cited Violations of 10 CFR 20.1902 and 10 CFR 20.1601 were identified for the failure to post and control access to high radiation areas. Corrective actions taken by the licensee for this finding included establishing control through postings and barricades. The cause of this finding is related to the cross-cutting element of human performance. (Section 2OS1).

Cornerstone: Public Radiation Safety

Green.

An inspector-identified finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified for the failure to perform surveys that are necessary to comply with the regulations in 10 CFR Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present prior to pumping liquids from blowdown line vacuum breaker valve vaults to the environment.

Specifically, the conditions found at 0CW276 (vault No. 6) on July 7, 2005, were outside the parameters of the original assessment, and the licensee did not evaluate the change of conditions for the potential radiological hazards to ensure compliance with 10 CFR 20.1301, which limits radiation exposure to a member of the public to 0.1 rem.

The issue was more than minor because the issue was associated with the Program/Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Since the releases were limited to licensee owned property, the licensee has not measured any licensed material beyond its property line, and the licensee's REMP has a monitoring well in the vicinity of the blowdown lines, the finding did not represent a failure to assess dose nor a failure to assess the environmental impact. Consequently, the fi nding was determined to be of very low safety significance. A Non-Cited Violation of 10 CFR 20.1501 was identified for the failure to make surveys to ensure compliance with 10 CFR 20.1301, which limits radiation exposure to a member of the public to 0.1 rem. Corrective actions taken by the licensee for this finding included performing surveys of the soil surrounding the vacuum breaker vault for radionuclides, establishing additional groundwater monitoring wells, sealing the vacuum breaker vaults, and inst alling of an automated leak detection system.

The cause of this finding is related to the cross-cutting element of problem identification and resolution. (Section 2PS1).

4

B.Licensee Identified Violations

Two violations of very low safety significance, which were identified by the licensee, have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and the licensee's corrective action tracking numbers are listed in

Section 4OA7 of this report.

5

REPORT DETAILS

Summary of Plant Status Unit 1 operated at or near full power throughout the first part of the inspection period. On August 25, 2006, the unit commenced a coast down for the upcoming refueling outage. On

September 10, 2006, at 11:00 p.m. the licensee opened the main generator output breaker and

entered a planned refueling outage. At the end of the report period the licensee was still in the

outage.Unit 2 operated at or near full power throughout the inspection period with the following exception: on July 29, 2006, Unit 2 power was reduced to approximately 63% due to offsite

transmission line issues. The unit was returned to full power the following day.1.REACTOR SAFETYCornerstone: Initiating Events, Mitigating Systems, Barrier Integrity andEmergency Preparedness

1R04 Equipment Alignment

Main article: IP 71111.04

.1Partial Walkdowns

a. Inspection Scope

The inspectors performed three partial walkdown samples of accessible portions of trains of risk-significant mitigating systems equipment during times when the trains were

of increased importance due to the redundant trains or other related equipment being

unavailable. The inspectors utilized the valve and electric breaker lineups and

applicable system drawings to determine that the components were properly positioned

and that support systems were lined up as needed. The inspectors also examined the

material condition of the components and observed operating parameters of equipment

to determine that there were no obvious deficiencies. The inspectors used the

information in the appropriate sections of the UFSAR and TS to determine the functional

requirements of the systems.

The inspectors verified the alignment of the following:

Unit 2 Emergency Core Cooling System;*Unit 1 Electrical Bus 131X; and

Unit 2 Train A Emergency Diesel Generator.

The inspectors also reviewed selected issues documented in IRs, to determine if they had been properly addressed in the licensee's corrective action program. The

documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

6 No findings of significance were identified..2Complete Walkdown

a. Inspection Scope

During the inspection, the inspectors finished one complete system alignment inspection of the accessible portions of the Spent Fuel Pool Cooling system after Unit 1 core was

offloaded. This system was selected because it was considered both safety-related, and

risk significant for the plant condition. The inspection consisted of the following

activities:*a review of plant procedures (including selected abnormal and emergency procedures), drawings, and the UFSAR to identify proper system alignment;*a review of outstanding work requests on the system;

a review of outstanding temporary modifications on the system;

a review of the system health information; and

a walkdown of the system to determine proper alignment, component accessibility, availability, and current condition.

The inspectors also reviewed selected issues documented in IRs, to determine if they had been properly addressed in the licensee's corrective actions program. The

documents reviewed during this inspection were listed in the Attachment at the end of

this report.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05).1Quarterly Walkdowns

a. Inspection Scope

The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of fire fighting equipment; the control of transient

combustibles and ignition sources; and on the condition and operating status of installed

fire barriers. The inspectors reviewed applicable portions of the Byron Station Fire

Protection Report and selected fire areas for inspection based on their overall

contribution to internal fire risk, as documented in the Individual Plant Examination of

External Events Report.

The inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were

unobstructed; that transient material loading was within the analyzed limits; and that fire

doors, dampers, and penetration seals appeared to be in satisfactory condition. The

Byron Station Pre-Fire Plans applicable for each area inspected were used by the

inspectors to determine approximate locations of firefighting equipment.

7 The inspectors completed eight inspection samples by examining the plant areas listed below to observe conditions related to fire protection:*Circulating Water Pump House (Zone 18.12-0);*Auxiliary Building Elevation 383' General Area (Zone 11.4-0);

Auxiliary Building Elevation 401' General Area (Zone 11.5-0);

Turbine Building 451' General Area (Zone 8.6-0);

Unit 2 Division 21 Miscellaneous Electrical Equipment and Battery Room (Zone 5.2-2);*Unit 2 Auxiliary Electrical Equipment Room (Zone 5.5-2);

Unit 1 Train B Diesel Generator & Day Tank Room (Zone 9.1-1 & 9.4-1); and

Unit 1 Containment Pipe Penetration Area (Zone 11.3-1).

The inspectors reviewed selected issues documented in CRs, to determine if they had been properly addressed in the licensee's corrective action program. The inspectors

also verified that minor issues identified during the inspection were entered into the

licensee's corrective action program. The documents reviewed during this inspection

are listed in the Attachment to this report.

b. Findings

Failure to Maintain Fire Barriers in Accordance with Fire Protection Program Introduction

The inspectors identified a Green finding and associated Non-Cited Violation of Byron Facility Operating License Nos. NPF-37 and NPF-66, Condition 2.c.6, for failing to maintain the firewall separating the Auxiliary Building from the penetration

area in accordance with the approved fire protection program.

Description

On September 25, 2006, during a routine fire protection walkdown of the Unit 1 auxiliary building mechanical penetration room, referred to by the licensee as

Area 5, the inspectors noted that there were ropes running through fire barriers in 5 inch

diameter conduit in cable tray 1757D C1E. The inspectors notified the licensee who

performed an independent walkdown and verified the installation of unapproved material

in the fire seals. This was inconsistent with Section 2.3.11.41 of the Fire Protection

Report, which described the fire area analysis for the 401 elevation of the Auxiliary

Building and stated that rated fire barriers separate this zone from the remainder of the

plant.The inspectors reviewed the Fire Protection Report and did not identify any existing deviations allowing for the existence of this condition. The licensee entered this issue in

their corrective action program for resolution (IR 536504) and implemented

compensatory actions that included hourly fire watches. By the end of the report period

the licensee had repaired the fire seals, restoring them to operable and then suspended

the compensatory actions.

Analysis:

The inspectors determined that the licensee's failure to maintain the fire seal between fire zone 11.3-1 and fire zone 11.5-0 in accordance with the approved fire

protection program was a performance deficiency warranting a significance

determination. Furthermore, the issue was considered more than minor because the 8 finding affected the attribute of protection against external factors (i.e. fire) of the Mitigating System Cornerstone. This finding was of very low safety significance because

the fire seals were of small diameter, traveled 45 feet with two 90 degree bends and the

location of combustibles were positioned such that the piping penetration end of the fire

seals would not be subject to direct flame impingement.

Enforcement

Byron Plant Operating License Condition 2.c.6 stated, in part, that "The licensee shall implement and maintain in effect all provisions of the approved fire

protection program as described in the UFSAR." Section 9.5.1 of the UFSAR stated that

"The design bases, system descriptions, safety evaluation, inspection and testing

requirements, personnel qualification, and training are described in Reference 1 [the Fire

Protection Report]." Contrary to the above, the licensee failed to maintain the firewall

separating the Auxiliary Building from the penetration area in accordance with the

approved fire protection program. Fire seals were required to be provided in this firewall, except where an evaluation had been perfo rmed and approved to allow a deviation.

Two sleeves containing fire seals had pulling ropes embedded in the fire seals in the

firewall separating the Auxiliary Building General Area 401 from the Unit 1 piping

penetration area; also, no evaluation or exemption existed to justify this configuration.

Because this issue was entered into the corrective action program as IR 536504, and

the finding was of very low safety significance, this violation is being treated as an NCV

consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000454/2006004-

01;05000455/2006004-01, Failure to Maintain Fire Barrier in Accordance with Fire

Protection Program).2Annual Drill Observation

a. Inspection Scope

The inspectors assessed the fire brigade performance and the drill evaluator's critique during a fire brigade drill conducted on August 31, 2006. The drill simulated a fire in the

Nuclear Station Work Permit building. The inspectors also observed an actual fire

brigade response to a fire alarm received in the Auxiliary Building on August 28, 2006.

Details of the fire response were documented in Section

4OA3 of this report.

The

documents reviewed for this portion of the inspection are listed in the Attachment to this

report.The inspectors focused on command and control of the fire brigade activities; fire fighting and communication practices; material condition and use of fire fighting

equipment; and implementation of pre-fire plan strategies. The inspectors evaluated the

fire brigade performance using the licensee's established fire drill performance

procedure criteria. An annual inspection sample was not completed in this report since

not all aspects of the inspectible areas were reviewed. Observation and evaluation of

other important drill activities designated in Section 02.02 of the inspection procedure

will be performed during subsequent observations of licensee drill activities.

b. Findings

No findings of significance were identified.

91R11Licensed Operator Requalification (71111.11).1Resident Inspector Quarterly Review

a. Inspection Scope

The inspectors completed one inspection sample by observing and evaluating an operating crew during a steam line break outside containment with failure of all MSIV's

to close. The inspectors evaluated crew performance in the areas of:*Clarity and formality of communications;*Ability to take timely actions;

Prioritization, interpretation, and verification of alarms;

Procedure use;

Control board manipulations;

Supervisor's command and control;

Management oversight; and

Group dynamics.

The inspectors verified that the crew completed the critical tasks listed in the above simulator guide. The inspectors also compared simulator configurations with actual

control board configurations. For any weaknesses identified, the inspectors observed

the licensee evaluators to determine whether they also noted the issues and discussed

them in the critique at the end of the session. The inspectors verified that minor issues

were placed into the licensee's corrective action program.

The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12).1Resident Inspector Quarter Review

a. Inspection Scope

The inspectors completed two inspection samples by evaluating the licensee's implementation of the maintenance rule, 10 CFR 50.65, as it pertained to identified

performance problems associated with the following structures, systems, and/or

components:*Unit 1 Train B Emergency Diesel Generator Trip During Cooldown Cycle; and*Unit 1 Molded Case Circuit Breaker Failures.

The inspectors evaluated the licensee's appropriate handling of structures, systems, and components (SSC) condition problems in terms of appropriate work practices and 10 characterizing reliability issues. Equipment problems were screened for review using a problem oriented approach. Work practices related to the reliability of equipment

maintenance were observed during the inspection period. Items chosen were risk

significant, and extent of condition was reviewed as applicable. Work practices were

reviewed for contribution to potential degraded conditions of the affected SSCs. Related

work activities were observed and corrective actions were discussed with licensee

personnel. The licensee's handling of the issues being reviewed was evaluated under

the requirements of the maintenance rule.

The inspectors also reviewed selected issues documented in CRs, to determine if they had been properly addressed in the licensee's corrective action program. The

documents reviewed during this inspection are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

The inspectors reviewed the licensee's management of plant risk during emergent maintenance activities or during activities where more than one significant system or

train was unavailable. The inspectors chose activities based on their potential to

increase the probability of an initiating event or impact the operation of safety-significant

equipment. The inspectors verified that the evaluation, planning, control, and

performance of the work were done in a manner to reduce the risk and the work duration

was minimized where practical. The inspecto rs also verified that contingency plans were in place where appropriate.

The inspectors reviewed configuration risk assessment records, UFSAR, TS, and Individual Plant Examination. The inspectors also observed operator turnovers, observed plan-of-the-day meetings, and reviewed other related documents to determine

that the equipment configurations had been properly listed, that protected equipment

had been identified and was being controlled where appropriate, and that significant

aspects of plant risk were being communicated to the necessary personnel.

The inspectors completed five inspection samples by reviewing the following activities:

Emergent Failure of the Unit 1 Train B Emergency Diesel Generator During Monthly Run;*Unit 2 Train B Residual Heat Removal Pump Out of Service (OOS) while Unit 1 East Main Power Transformer Sensor was Replaced and while Unit 0 G train SX

Cooling Tower was OOS;*Unit 1 Train A Essential Service Water Pump Work Window while Unit 0 Essential Service Water Makeup Pump was OOS;*Emergent Schedule Change Due to Thunderstorms and the Delay in the Flood-up of the Reactor Vessel for Unit 1 with a Consequential Increase in Loss of Core

Cooling While at Reduced Inventory; and 11*Unit 2 Train A Direct Current (DC) Bus Cross-tied to Unit 1 Train A while Unit 1 Train A Emergency Diesel Generator was OOS.

The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors evaluated plant conditions, selected condition reports, engineering evaluations, and operability determinations for risk-significant components and systems

in which operability issues were questioned. These conditions were evaluated to

determine whether the operability of components was justified.

The inspectors completed eight inspection samples by reviewing the following evaluations and issues:*Unit 1 Train A Diesel Generator Excessive Combustion Air Moisture;*Seismic Monitoring System Ou t of Service and Impact Upon EALs;*Unit 1 Pressurizer Power Operated Relief Valve Elevated Tailpipe Temperature;

Unit 1 Train A Containment Spray Additive System Valve Weld Leak;

Unit 2 Train B Essential Service Water Pump Bearing Slinger Ring Not Rotating;

Unit 2 Train B Auxiliary Feedwater Pump Jacket Water Expansion Tank Leak;

Unit 1 Molded Case Circuit Breaker Failure; and

Unit 1 Emergency Core Cooling System Piping Air Entrainment.

The inspectors compared the operability and design criteria in the appropriate section of the TS including the TS Basis, the Technical Requirements Manual (TRM) and UFSAR

to the licensee's evaluations to determine that the components or systems were

operable. The inspectors determined whether compensatory measures, if needed, were

taken, and determined whether the evaluations were consistent with the requirements of

licensee procedures. The inspectors also discussed the details of the evaluations with

the shift managers and appropriate members of the licensee's engineering staff.

The inspectors also reviewed selected issues documented in IRs, to determine if they had been properly addressed in the licensee's corrective action program. The

documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed the post maintenance testing activities associated with maintenance or modification of mitigating, barrier integrity, and support systems that

were identified as risk significant in the licensee's risk analysis. The inspectors reviewed

these activities to determine that the post maintenance testing was performed

adequately, demonstrated that the maintenance was successful, and that operability was

restored. During this inspection activity, the inspectors interviewed maintenance and

engineering department personnel and reviewed the completed post maintenance

testing documentation. The inspectors used the appropriate sections of the TS, TRM, and UFSAR, and other related documents to evaluate this area.

The inspectors completed six inspection samples by observing and evaluating the post maintenance testing subsequent to the following maintenance activities:*Unit 1 Train A Safety Injection to Charging Pump Suction Header Cross-Tie Isolation Valve Actuator Rebuild;*Unit 1 Train A Emergency Diesel Generator valve work;

Unit 1 Train A Essential Services Water Pump oil cooler inspection;

Unit 1 Train B Centrifugal Charging Pump Work Window;

Unit 0 Train B Essential Service Water Makeup Pump Work Window; and

Unit 1 Division 11 DC Bus Circuit Breaker Change-out.

The inspectors also reviewed selected issues documented in CR's to determine if they had been properly addressed in the licensee's corrective action program. The

documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20)

a. Inspection Scope

The inspectors observed the licensee's performance during Refueling Outage B1R14 beginning September 10, 2006. As of September 30, 2006, the licensee has not

finished all the outage activities. Therefore the inspection sample was not competed for

this report.

The inspectors evaluated the licensee's conduct of refueling outage activities to assess the licensee's control of plant configuration and management of shutdown risk. The

inspectors reviewed configuration management to verify that the licensee maintained

defense-in-depth commensurate with the shutdown risk plan; reviewed major outage

work activities to ensure that correct system lineups were maintained for key mitigating

systems; and observed refueling activities to verify that fuel handling operations were

performed in accordance with the TS, TRM, UFSAR and approved procedures. The

inspectors interviewed operations, engineering, work control, radiological protection, and

maintenance department personnel during their inspection activities. The inspectors 13 also attended outage-related status and pre-job briefings as well as Radiation Protection ALARA [As Low As Reasonably Achievable] briefings. Other major outage activities

evaluated included evaluating the licensee's control of:*containment penetrations in accordance with the TS;*structures, systems or components (SSCs) which could cause unexpected reactivity changes;*flow paths, configurations, and alternate means for reactor coolant system inventory addition;*SSCs which could cause a loss of inventory;

RCS pressure, level, and temperature instrumentation;

spent fuel pool cooling during and after core offload;

switchyard activities and the configuration of electrical power systems in accordance with the TS and shutdown risk plan; and*SSCs required for decay heat removal.

The inspectors observed portions of the plant cooldown, including the transition to shutdown cooling, to verify that the licensee controlled the plant cooldown in accordance

with the TS. In addition, the inspectors completed numerous visual inspections inside

the Unit 1 containment. This included a tour of the Unit 1 containment at Mode 3 during

the cooldown at the beginning of B1R14 so that the inspectors could assess the initial

material condition of equipment inside containment immediately following the operating

cycle. During the visual inspections the inspectors focused on the material condition of

the equipment and particularly on any indication of boric acid.

In addition, the inspectors evaluated portions of the restart preparation activities to verify that requirements of the TS and administrative procedure requirements were met prior to

changing operational modes or plant configurations. Major restart preparation

inspection activities performed included:*verification that core reload was completed in accordance with the core loading plan for Byron Unit 1 Cycle 14;*evaluation of foreign material exclusion control practices during significant work activities;*verification that correct system lineups were maintained for key mitigatingsystems; and*inspection of the containment building to assess material condition and search for loose debris, which if present, could be transported to the containment

recirculation sumps and cause restriction of flow to the emergency core cooling

system pump suctions during loss-of-coolant accident conditions.

The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

141R22Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors witnessed selected surveillance tests and/or reviewed test data to determine that the equipment tested using the surveillance procedures met the TS, the

TRM, the UFSAR and licensee procedural requirements. The inspectors also reviewed

applicable design documents including plant drawings, to verify that the surveillance

tests demonstrated that the equipment was capable of performing its intended safety

functions. The activities were selected based on their importance in ensuring mitigating

systems capability and barrier integrity.

These activities represented six routine samples. The following surveillance tests were selected:*Unit 2 Train A Emergency Diesel Generator Monthly Operability Run;*Unit 1 Reactor Containment Fan Cooler Monthly Surveillance;

Unit 2 Train B Solid State Protection System Bi-Monthly Surveillance;

Unit 1 Simultaneous Start of Both Auxiliary Feedwater Pumps;

Unit 1 Train B Diesel Generator Safety Injection Sequencer Test; and

Unit 1 Division 11 A Train 125V Battery Bank Service Test.

Additionally the inspectors used the documents listed in the attachment to this report to determine that the testing met the frequency requirements; that the tests were

conducted in accordance with procedures, that the test acceptance criteria were met;

and that the results of the tests were properly reviewed and recorded. The inspectors

verified that the individuals performing the tests were qualified to perform the test in

accordance with the licensee's requirements, and that the test equipment used during

the test were calibrated within the specified periodicity. In addition, the inspectors

interviewed operations, maintenance, and engineering department personnel regarding

the tests and test results. The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.1EP6Drill Evaluation (71114.06)

a. Inspection Scope

On August 22, 2006, the inspectors complete one inspection sample by observing an Out of the Box Operator Requalification training that had emergency preparedness

exercise aspects. The inspectors assessed the licensee's exercise performance and

looked for weaknesses in the risk significance areas of emergency classification, notification and protective action development. The inspectors observed the licensee's

performance from the simulator control room. The inspectors compared issues noted 15 during their observations to those identified during the licensee's critique. Additionally, the inspectors verified that items identified during the licensee's critique were

appropriately entered into their corrective action program.

The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.2.RADIATION SAFETY

===Cornerstone: Occupational Radiation Safety2OS1Access Control to Radiologically Significant Areas (IP 71121.01).1Plant Walkdowns and Radiation Work Permit Reviews

a. Inspection Scope

=

The inspectors reviewed licensee controls and surveys in the following three radiologically significant work areas within radiation areas, high radiation areas and

airborne radioactivity areas in the plant and reviewed work packages which included

associated licensee controls and surveys of these areas to determine if radiological

controls including surveys, postings and barricades were acceptable: *Emergency Core Cooling System (ECCS) Sump Modification;*Pressurizer Weld Overlay; and

Reactor Disassembly.

This review represented one inspection sample.

The inspectors reviewed the radiation work permits (RWPs) and work packages used to access these three areas and other high radiation work areas to identify the work control

instructions and control barriers that had been specified. Electronic dosimeter alarm set

points for both integrated dose and dose rate were evaluated for conformity with survey

indications and plant policy. Workers were interviewed to verify that they were aware of

the actions required when their electronic dosimeters noticeably malfunctioned or

alarmed. This review represented one inspection sample.

The inspectors walked down and surveyed (using an NRC survey meter) these three areas to verify that the prescribed RWP, procedure, and engineering controls were in

place; that licensee surveys and postings were complete and accurate; and that air

samplers were properly located. This review represented one inspection sample.

The inspectors reviewed RWPs for potential airborne radioactivity areas to verify barrier integrity and engineering controls performance (e.g., HEPA ventilation system

operation) and to determine if there was a potential for individual worker internal 16 exposures of greater than 50 millirem committed effective dose equivalent. There were no areas where there was a potential for individual worker internal exposures of greater

than 50 millirem committed effective dose equivalent. Work areas having a history of, or

the potential for, airborne transuranics were evaluated to verify that the licensee had

considered the potential for transuranic isotopes and provided appropriate worker

protection. There where no areas having a history of, or the potential for, airborne

transuranics. This review represented one inspection sample.

The adequacy of the licensee's internal dose assessment process for any actual internal exposures greater than 50 millirem committed effective dose equivalent was assessed.

There were no internal exposures greater than 50 millirem committed effective dose

equivalent. This review represented one inspection sample.

b. Findings

Introduction

A finding of very low safety significance and two associated Non-Cited Violations of NRC requirements were identified for the failure to post and control access

to a high radiation area.

Description

A walkdown of the facility was conducted by the NRC inspectors and a member of the licensee's staff during the afternoon of September 11, 2006. This

walkdown identified that an opening of approximately 3'x4' in size that led to a pipe

chase next to the Unit 1 Filter Valve Aisle located on the 383' elevation of the Auxiliary

Building. This opening was created when a portion of a block shield wall was removed

to support a specific job evolution in the Unit 1 Filter Valve Aisle. The inspectors

questioned the licensee staff as to the systems that lie within the pipe chase that was

rendered accessible by removing the blocks. Based upon the answer provided and the

upcoming scheduled plant evolution of forced oxidation, the inspectors questioned the

characterization of the pipe chase and the adequacy of the posting and controls for the

current conditions or the expected conditions over the next few hours.

The forced oxidation process is performed to reduce or remove radioactive source term from the primary coolant system and, therefore, to lower the personnel dose that is

accumulated during the refueling outage. The process introduces a chemical to the

reactor coolant system which loosens material that has plated out inside the system.

This material is removed from the system via plant demineralizers and through filtration, usually within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The licensee implemented controls for areas that had been

defined based upon previous plant experiences. The block shield wall was intact, different than current configuration, during previous evolutions of forced oxidation.

The licensee completed the survey of this area at 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months on September 13, 2006, approximately 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months after initiating the forced oxidation process. That survey

identified dose rates in excess of 100 mR/hour at the plane of the opening. These

conditions met the definition as a High Radiation Area, but the area was not posted or

controlled as a High Radiation Area.

Analysis:

The failure to post and control access to High Radiation Areas represents a performance deficiency as defined in NRC Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening." The inspectors 17 determined that the issue was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to

ensure the adequate protection of the worker health and safety from exposure to

radiation from radioactive material during routine civilian nuclear reactor operation.

Therefore, the issue was more than minor and represented a finding which was

evaluated using the Significance Determination Process (SDP).

Since the finding involved the ability to protect workers from exposure to radiation, the inspectors utilized IMC 0609, Appendix C, "Occupational Radiation Safety SDP," to

assess its significance. The inspectors determined that the finding did not concern

unintended collective dose resulting from a deficiency in the ALARA planning or work

control or exposure control. The inspectors also determined that the finding did not

involve an overexposure or the substant ial potential for an overexposure. The inspectors determined that the finding did not compromise the licensee's ability to

assess dose. Consequently, the inspectors concluded that the SDP assessment for this

finding was of very low safety significance (Green).

As described above, the removal of the ~ 3'x4' block wall was authorized to support a work activity. However, the evaluation that provided this authorization was not sufficient

to provide limits, controls, or compensatory actions for subsequent plant evolutions.

Consequently, this deficiency has a cross cutting aspect for Human Performance.

Specifically, the licensee did not use a systematic decision making process and did not

obtain interdisciplinary input on a risk-significant decision.

Enforcement

10 CFR 20.1902 requires the licensee to post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "CAUTION, HIGH RADIATION AREA" or "DANGER, HIGH RADIATION AREA. Additionally, 10

CFR 20.1601 specifies the requirements for control of access to high radiation areas.

As provided in Technical Specification 5.7.1, the licensee is authorized to implement

alternate controls to those stated in 10 CFR 20.1601 for areas that do not exceed 1000

millirem per hour.

Contrary to the above, as of 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months on September 13, 2006, the entrance to the Unit 1 Filter Valve Aisle located on the 383' elevation of the Auxiliary Building, a high

radiation area with a radiation dose rate of approximately 135 millirem in one hour at

plane of the penetration, was not posted or controlled by any of the methods described

in 10 CFR 20.1902, 10 CFR 20.1601, or Technical Specification 5.7.1.

Corrective actions taken by the licensee included making the pipe chase inaccessible by bolting a plate over the opening and placing information postings over the bolted plate.

In addition, the licensee revised the posting and controls within the Unit 1 Filter Valve

Aisle to that of a high radiation area. Since the licensee documented this issue in its

corrective action program (IR 531013) and because the violations are of very low safety

significance, they are being treated as Non-Cited Violations (NCV 05000454/2006004-02; 05000455/2006004-02).

.2 Problem Identification and Resolution

a. Inspection Scope

18 The inspectors reviewed three corrective action reports related to access controls and high radiation area radiological incidents. Staff members were interviewed and

corrective action documents were reviewed to verify that follow-up activities were being

conducted in an effective and timely manner commensurate with their importance to

safety and risk based on the following:*initial problem identification, characterization, and tracking;*disposition of operability/reportability issues;

evaluation of safety significance/risk and priority for resolution;

identification of repetitive problems;

identification of contributing causes;

identification and implementation of effective corrective actions;

resolution of Non-Cited Violations (NCVs) tracked in the corrective action system; and*implementation/consideration of risk significant operational experience feedback.

This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Job-In-Progress Reviews

a. Inspection Scope

The inspectors observed the following three jobs that were being performed in radiation areas, airborne radioactivity areas, or high radiation areas for observation of work

activities that presented the greatest radiological risk to workers: *ECCS Sump Modification;*Pressurizer Weld Overlay; and

Reactor Disassembly.

The inspectors reviewed radiological job requirements for these three activities including RWP requirements and work procedure requirements, and attended As-Low-As-Is-

Reasonably-Achievable (ALARA) job briefings. This review represented one inspection sample. Job performance was observed with respect to these requirements to verify that radiological conditions in the work area were adequately communicated to workers

through pre-job briefings and postings. The inspectors also verified the adequacy of

radiological controls including required radiation, contamination, and airborne surveys

for system breaches; radiation protection job coverage which included audio and visual

surveillance for remote job coverage; and contamination controls. This review

represented one inspection sample.

b. Findings

19 No findings of significance were identified.

.4 Radiation Worker Performance

a. Inspection Scope

During job performance observations, the inspectors evaluated radiation worker performance with respect to stated radiation protection work requirements and evaluated

whether workers were aware of the significant radiological conditions in their workplace, the RWP controls and limits in place, and that their performance had accounted for the

level of radiological hazards present. This review represented one inspection sample.

The inspectors reviewed radiological problem reports which found that the cause of the event was due to radiation worker errors to determine if there was an observable pattern

traceable to a similar cause and to determine if this perspective matched the corrective

action approach taken by the licensee to resolve the reported problems. These

problems, along with planned and taken corrective actions were discussed with the

Radiation Protection Manager. This review represented one inspection sample.

b. Findings

No findings of significance were identified..5Radiation Protection Technician (RPT) Proficiency

a. Inspection Scope

During job performance observations, the inspectors evaluated RPT performance with respect to radiation protection work requirements and evaluated whether they were

aware of the radiological conditions in their workplace, the RWP controls and limits in

place, and if their performance was consistent with their training and qualifications with

respect to the radiological hazards and work activities. This review represented one

inspection sample.

The inspectors reviewed radiological problem reports which, found that the cause of the event was radiation protection technician error, to determine if there was an observable

pattern traceable to a similar cause and to determine if this perspective matched the

corrective action approach taken by the licensee to resolve the reported problems. This

review represented one inspection sample.

b. Findings

No findings of significance were identified.2OS2As-Low-As-Is-Reasonably-Achievable Planning and Controls (ALARA) (IP 71121.02).1

Inspection Planning

a. Inspection Scope

20 The inspectors reviewed plant collective exposure history, current exposure trends, ongoing and planned activities in order to assess current performance and exposure

challenges. This included determining the plant's current 3-year rolling average for

collective exposure in order to help establish resource allocations and to provide a

perspective of significance for any resulting inspection finding assessment. This review

represented one inspection sample.

The inspectors reviewed the outage work scheduled during the inspection period and associated work activity exposure estimates fo r the following three work activities which were likely to result in the highest personnel collective exposures: *Emergency Core Cooling System (ECCS) Sump Modification;*Pressurizer Weld Overlay; and

Reactor Disassembly.

This review represented one inspection sample.

The inspectors reviewed site specific trends in collective exposures and source-term measurements. The inspectors reviewed procedures associated with maintaining

occupational exposures ALARA and processes used to estimate and track work activity

specific exposures. This review represented two inspection samples.

b. Findings

No findings of significance were identified..2Radiological Work Planning

a. Inspection Scope

The inspectors evaluated the licensee's list of planned work activities for Unit 1 Refueling Outage 14 ranked by estimated exposure that were in progress and reviewed

the following three work activities of exposure significance: *Emergency Core Cooling System (ECCS) Sump Modification;*Pressurizer Weld Overlay; and

Reactor Disassembly.

This review represented one inspection sample.

For these three activities, the inspectors re viewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements in order to verify that the

licensee had established procedures and engineering and work controls that were based

on sound radiation protection principles in order to achieve occupational exposures that

were ALARA. This also involved determining if the licensee had reasonably grouped the

radiological work into work activities, based on historical precedence, industry norms, and/or special circumstances. This review represented one inspection sample.

The inspectors compared the results achieved, including dose rate reductions and 21 person-rem used, with the intended dose established in the licensee's ALARA planning for these three work activities. Reasons for inconsistencies between intended and

actual work activity doses were review ed. This review represented one inspectionsample.

b. Findings

No findings of significance were identified..3Verification of Dose Estimates and Exposure Tracking Systems

a. Inspection Scope

The licensee's process for adjusting exposure estimates or re-planning work, when unexpected changes in scope, emergent work or higher than anticipated radiation levels

were encountered, was evaluated. This included determining whether adjustments to

estimated exposure (intended dose) were based on sound radiation protection and

ALARA principles and not adjusted to account for failures to control the work. The

frequency of these adjustments was reviewed to evaluate the adequacy of the original

ALARA planning process. This review represented one inspection sample.

b. Findings

No findings of significance were identified..4Job Site Inspections and ALARA Control

a. Inspection Scope

The inspectors observed the following three jobs that were being performed in radiation areas, airborne radioactivity areas, or high radiation areas for observation of work

activities that presented the greatest radiological risk to workers.*Emergency Core Cooling System (ECCS) Sump Modification;*Pressurizer Weld Overlay; and

Reactor Disassembly.

The licensee's use of engineering controls to achieve dose reductions was evaluated to verify that procedures and controls were consistent with the licensee's ALARA reviews, that sufficient shielding of radiation sources was provided, and that the dose expended

to install/remove the shielding did not exceed the dose reduction benefits afforded by the

shielding. This review represented one inspection sample.

b. Findings

No findings of significance were identified.

.5 Radiation Worker Performance

a. Inspection Scope

Radiation worker and RPT performance was observed during work activities being performed in radiation areas, airborne radioactivity areas, and high radiation areas that

presented the greatest radiological risk to workers. The inspectors evaluated whether

workers demonstrated the ALARA philosophy in practice by being familiar with the work

activity scope and tools to be used, by utilizing ALARA low dose waiting areas and that

work activity controls were being complied with. Also, radiation worker training and skill

levels were reviewed to determine if they were sufficient relative to the radiological

hazards and the work involved. This review represented one inspection sample.

b. Findings

No findings of significance were identified.2OS3Radiation Monitoring Instrumentation and Protective Equipment (71121.03).1

Inspection Planning

a. Inspection Scope

The inspectors reviewed the Byron Station Updated Final Safety Analysis Report (UFSAR) to identify applicable radiation monitors associated with measuring transient

high and very high radiation areas including those used in remote emergency

assessment. The inspectors identified the types of portable radiation detection

instrumentation used for job coverage of high radiation area work including instruments

used for underwater surveys, fixed area radiat ion monitors used to provide radiological information in various plant areas, and continuous air monitors used to assess airborne

radiological conditions and work areas with the potential for workers to receive a

50 millirem or greater committed effective dose equivalent (CEDE). Contamination monitors, whole body counters, and those radiation detection instruments utilized for the

release of personnel and equipment from the radiologically controlled area (RCA) were

also identified.

These reviews represented two inspection samples.

b. Findings

No findings of significance were identified..2Walkdowns of Radiation Monitoring Instrumentation

a. Inspection Scope

The inspectors conducted walkdowns of selected area radiation monitors (ARMs)in the Unit 1 and 2 Auxiliary Buildings to verify that they were located as described in

the UFSAR and were adequately positioned relative to the potential source(s) of

radiation they were intended to monitor. Walkdowns were also conducted of those

areas where portable survey instruments were calibrated/repaired and maintained for 23 radiation protection (RP) staff use to determine if those instruments designated "ready for use" were sufficient in number to support the radiation protection program, had

current calibration stickers, were operable, and were in adequate physical condition.

Additionally, the inspectors observed the licensee's instrument calibration units and the

radiation sources used for instrument checks to assess their material condition and

discussed their use with RP staff to determine if they were used appropriately. Licensee

personnel demonstrated the methods for performing source checks of portable survey

instruments and for source checking personnel contamination and portal monitors used

at the egress to the RCA.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Calibration and Testing of Radiation Monitoring Instrumentation

a. Inspection Scope

Portable survey instrument calibrations were performed at an offsite Exelon facility.

Licensee personnel were observed performing source checks of selected instruments.

This included observing detector evaluation with check sources to determine if station

requirements were met. The inspectors reviewed records of calibration, operability, and

alarm setpoints of selected instruments and personnel monitoring devices. This review

included, but was not limited to the following:*Certificate of Calibration for Eberline Radiation Detection Device Model ASP-1/AC3-7, Serial No. 652/724197;*Certificate of Calibration for Eberline Radiation Detection Device Model E-530, Serial No. 1337; *Certificate of Calibration for Eberline Radiation Detection Device Model RM-14, Serial Nos. 7382 and 7528;*Calibration of Nuclear Enterprises Small Articles Monitor (SAM), Serial No. 478;

Units 1 and 2 High Range Containment Radiation Monitors; and

Auxiliary Building Vent Stack Wide Range Gas Radiation Monitor.

The inspectors evaluated those actions that would be taken when, during calibration or source checks, an instrument was found to be out of calibration by more than

50 percent. Those actions included an investigation of the instrument's previous usages

and the possible consequences of that usage since the last calibration or source check.

The inspectors also reviewed the licensee's 10 CFR Part 61 source term analyses to

determine if the calibration sources used were representative of the plant source term.

This review represented one sample.

b. Findings

No findings of significance were identified.

24.4Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed the licensee's self-assessments, audits, and condition reports that involved personnel contamination monitor alarms due to personnel internal

exposures to determine if identified problems were entered into the corrective action

program for resolution. There were no internal exposure occurrences greater than

50 millirem committed effective dose equivalent that were evaluated during the

inspection. However, the licensee's process for investigating this type of occurrence

was reviewed to determine if the affected personnel would be properly monitored

utilizing the appropriate equipment and if the data would be analyzed and internal

exposures properly assessed in accordance with licensee procedures. This review

represented one sample.

The inspectors reviewed corrective action program reports related to exposure of significant radiological incidents that involved radiation monitoring instrument

deficiencies since the last inspection in this area. Staff members were interviewed and

corrective action documents were reviewed to determine if follow-up activities were

being conducted in an effective and timely m anner commensurate with its importance to safety and risk based on the following:*Initial problem identification, characterization, and tracking;*Disposition of operability/reportability issues;

Evaluation of safety significance/risk and priority for resolution;

Identification of repetitive problems;

Identification of contributing causes;

Identification and implementation of effective corrective actions;

Resolution of Non-Cited Violations tracked in the corrective action system; and

Implementation/consideration of risk significant operational experience feedback.

This review represented one sample.

The inspectors evaluated the licensee's self-assessment activities to determine if they would identify and address repetitive deficiencies or significant individual deficiencies

observed in problem identification and resolution. This review represented one sample.

b. Findings

No findings of significance were identified.5Radiation Protection Technician Instrument Use

a. Inspection Scope

The inspectors determined if the calibration expiration and source response check data records on radiation detection instruments staged for use were current and observed

radiation protection technicians for appropriate instrument selection and self-verification

of instrument operability prior to use. This review represented one sample.

b. Findings

No findings of significance were identified..6Self-Contained Breathing Apparatus (SCBA) Maintenance/Inspection and User Training

a. Inspection Scope

The inspectors reviewed the status, maintenance and surveillance records of selected self-contained breathing apparatuses staged and ready for use in the plant and

assessed the licensee's capability for refilling and transporting self-contained breathing

apparatus air bottles to and from the control room during emergency conditions. The

inspectors determined whether control room operators and other emergency response

and radiation protection personnel were trained and qualified in the use of self-contained

breathing apparatuses including personal bottle change-out. The inspectors also

reviewed the training and qualification records for selected individuals on each control

room shift crew and selected individuals from each designated department that were

currently assigned emergency duties, including onsite search and rescue. This review

represented one sample.

The inspectors reviewed the self-contained breathing apparatus manufacturer's maintenance training certifications for licensee personnel qualified to perform self-

contained breathing apparatus maintenance on vital components (regulator and low

pressure alarm). The inspectors reviewed maintenance records for several self-

contained breathing apparatuses designated as "ready for service." The inspectors

verified that maintenance was performed by qualified personnel over the past five years.

The inspectors also determined if the required, periodic air cylinder hydrostatic testing

was current and documented. The inspectors also evaluated if the licensee's

maintenance procedures were consistent with the self-contained breathing apparatus

manufacturer's maintenance manuals. This review represented one sample.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety

2PS1Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (71122.01).1Integrity of the Circulating Water Blowdown Line

a. Inspection Scope

The inspectors reviewed the licensee's evaluation associated with Unresolved Item (URI) 050000454/2006002-02; 050000455/2006002-02 regarding the licensee's lack of

a circulating water blowdown line pipe integrity verification. Industry experience had

shown that the failure of circulating water blow-down line vacuum breakers resulted in

the release of contaminated water into the offsite environment resulting in groundwater

contamination. On February 2, 2006, the licensee could not demonstrate the integrity of 26 circulating water blowdown line vacuum breakers. Leakage of water from the blow-down line could result in the release of radioactive material into the environment via a release

path that was undefined in the ODCM and had no offsite dose estimates, and as

documented as an URI.

The inspectors reviewed the licensee's radiological assessment of leaks from the circulating water (CW) blowdown line that was identified in February 2006. The

inspectors reviewed historical records to evaluate the licensee's response to the leaks, including radiological surveys, dose assessments, and mitigative actions. The

inspectors' evaluation was performed to determine if the licensee adequately

implemented the requirements contained in 10 CFR Part 20 and the licensee's Technical

Specifications. The inspectors also reviewed:*Radiation protection surveys for affected areas;*Maintenance work orders for selected vacuum breaker valves associated with the releases;*Identification of potential pathways based upon release location;

Reports contained in the licensee's corrective action program for these events;

Parameters and results of licensee's groundwater characterization study;

Files that contain environmental contamination events; and

Select annual effluent release reports.

Additionally, the inspector's evaluated the licensee's corrective actions that included:

Sealing the vacuum breaker valve vaults (via grout and a waterproofing application);*Performing inspection and maintenance on the vacuum breaker valves;

Installation of additional groundwater monitoring wells;

Visual observations of each vacuum breaker valve vault during radioactive releases to identify leakage (compensatory action); and *Installation of an automated leak detection system.

b. Findings

Introduction

A finding of very low safety significance and an associated non-cited violation of NRC requirements were identified for the failure to survey water for

radioactive materials before releasing to the ground.

Description

The licensee began formally inspecting vacuum breaker valves in 1999 as result of a failure of similar vacuum breaker valves at the Braidwood facility. The

licensee attributed the presence of small amounts of water found in the vacuum breaker

vaults to be ground water that infiltrated the through a drain in the vault or from rain

water that seeped under the vault cover. On July 7, 2005, the vault containing valve

0CW276 (vault No. 6) was filled with water, and there were other indications that the

source was not ground water infiltration. Despite these indications, the vault was

emptied by pumping the water to ground. The licensee did not assess the source of the

water nor did the licensee perform a radiological survey prior to pumping the water to the

ground. The licensee recognized the potential for this water to contain radioactive

material from the liquid radioactive waste program when the failed valve was entered 27 into the corrective action program. Subsequently, the licensee performed an investigation that included sampling the soil around the vault for the presence of

radionuclides. The investigation was limited to analyzing the soil surrounding the

affected vault, as the area did not have free standing water. This analysis was

completed within two weeks from the time the condition was identified and determined

that the area was free from gamma emitting radionuclides.

During an NRC baseline inspection in the area of Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems in February 2006, the inspectors questioned

the licensee's evaluation of water that exist ed in the vacuum breaker valve vaults and the lack of radiological measurements of that water. In response to these questions, the

licensee then identified standing water in 5 of the 6 vaults with tritium concentrations

from 1,000 picocuries per liter to 80,000 picocuries per liter. Based on these sample

results, the licensee installed ground water monitoring wells near the six vacuum breaker

valve vaults along the station's discharge pipe to allow further sampling for tritium. Wells

beside four of the six vaults showed no detectable levels of tritium. Test wells beside the

other two showed low levels of tritium. One showed a concentration of about

3800 picocuries per liter, the other about 450 picocuries per liter. The licensee's

radiological environmental monitoring pr ogram (Rev. 2, 2002) included waterborne sampling/analyses from wells located 0.7 mile s, 1.0 miles, and 1.8 miles from the plant site and in the general area of the blowdown line. The licensee's results from these

wells as well as the 6 newly installed wells did not identified any detectable tritium

migration beyond the licensee's property.

Although the licensee recognized in July 2005 that water was apparently leaking from the valve and recognized the potential that leakage from a blowdown line vacuum

breaker could represent an unevaluated condition, the licensee did not take actions

necessary to evaluate the radiological hazards associated with leakage. Furthermore, the corrective action evaluation performed by the licensee in August 2005 did not

address or modify plant procedures or work order packages for inspection of vacuum

breakers to prevent further pumping water from the vaults to ground with sampling.

Analysis:

The failure to evaluate the potential radiological hazard associated with the leakage of water from the vacuum breaker valves and the subsequent discharge of that

water to the ground represents a performance deficiency as defined in NRC Inspection

Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue

Screening." The inspectors determined that the issue was associated with the

Program/Process attribute of the Public Radiation Safety Cornerstone and affected the

cornerstone objective to ensure adequate protection of public health and safety from

exposure to radioactive materials released into the public domain as a result of routine

civilian nuclear reactor operation. Therefore, the issue was more than minor and

represented a finding which was evaluated using the Significance Determination

Process (SDP).

Since the finding involved the ability to assess dose from radioactive effluents and maintain radiation doses to a member of the public within Appendix I design objectives, the inspectors utilized IMC 0609, Appendix D, "Public Radiation Safety SDP," to assess

its significance. The inspectors determined that the finding did not involve Radioactive

Material Control. Since the release did not migrate off the licensee's property, the 28 inspectors utilized the Effluent Release Program branch of the SDP. Although the licensee analyzed the surrounding soil for the release of gamma emitting radionuclides, tritium was not included in the analysis because the liquid content was no longer

available after pumping the vault. Tritium is the most predominant radionuclide in a

typical liquid radioactive waste release. Therefore, this limitation impaired the licensee's

ability to assess dose. The licensee's current sampling and evaluation did not indicate

any measurable release of radioactive material beyond the licensee's property. These

results indicated that the assessment of off site dose was not warranted; therefore, the

licensee did not fail to assess dose to the public. Consequently, the inspectors

concluded that the SDP assessment for this finding was of very low safety significance (Green).As described above, the licensee's August 2005 evaluation was limited to the water that was pumped out of vault No. 6 on July 7, 2005. The inspectors identified that the

evaluation did not review or change the practice of pumping water directly from the vault at any other vacuum breaker. Consequently, this deficiency has a cross cutting aspect

for Problem Identification and Resolution. Specifically, the corrective action evaluation

performed by the licensee in August 2005 did not address or modify plant procedures or work order packages for inspection of vacuum breakers to prevent pumping water from

the vaults to ground with sampling.

Enforcement

10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in

10 CFR Part 20 and that are reasonable under the circumstances to evaluate the extent

of radiation levels, concentrations or quantities of radioactive materials, and the potential

radiological hazards that could be present.

Pursuant to 10 CFR 20.1003, survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or

presence of radioactive material or other sources of radiation.

Contrary to the above, as of February 8, 2006, the licensee did not make adequate surveys to assure compliance with 10 CFR 20.1301, which limits radiation exposure to a

member of the public to 0.1 rem. Specifically, the conditions found at valve 0CW276 (vault No. 6) on July 7, 2005, were outside the parameters of the licensee's original

assessment, and the licensee did not evaluate the change of conditions for the potential

radiological hazards. A review of historical records indicated other occurrences of

pumping water from the vacuum breaker vaults to the ground without performing a survey.Corrective actions taken by the licensee for this finding included performing surveys of the soil surrounding the vacuum breaker vault, establishing additional groundwater

monitoring wells, sealing the vacuum breaker vaults, and installing an automated leak

detection system. Since the licensee documented this issue in its corrective action

program (AR 350931 and subsequent Apparent Cause Evaluation No. 478372) and

because the violation is of very low safety significance, it is being treated as a Non-Cited

Violation (NCV 05000454/2006004-03; 05000455/2006004-03). The associated URI is

closed.

292PS2Radioactive Material Processing and Transportation (71122.02).1Radioactive Waste System

Inspection Planning

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system description in the UFSAR for information on the types and amounts of radioactive waste (radwaste)

generated and disposed. The inspectors reviewed the scope of the licensee's audit

program with regard to radioactive material processing and transportation programs to

b. Findings

No findings of significance were identified..2Walkdown of Radioactive Waste Systems

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system description in the UFSAR and the most recent information regarding the types and amounts of radioactive

waste generated and disposed. The inspectors performed walkdowns of the liquid and

solid radwaste processing systems to ve rify that the systems agreed with the descriptions in the Updated Safety Analysis Report and the Process Control Program

and to assess the material condition and operability of the systems. The inspectors

reviewed changes to the waste processing system to verify the changes were reviewed

and documented in accordance with 10 CFR 50.59 and to assess the impact of the

changes on radiation dose to members of the public.

The inspectors reviewed the current processes for transferring waste resins into transportation containers to determine if appropriate waste stream mixing and/or

sampling procedures were utilized. The inspectors also reviewed the methodologies for

waste concentration averaging to determine if representative samples of the waste

product were provided for the purposes of waste classification in accordance with

10 CFR 61.55. During this inspection, the licensee was not conducting waste

processing. This review represented one sample.

b. Findings

No findings of significance were identified..3Waste Characterization and Classification

a. Inspection Scope

The inspectors reviewed the licensee's radiochemical sample analysis results for each of the licensee's waste streams, including dry active waste, resins, and filters. The 30 inspectors also reviewed the licensee's use of scaling factors to quantify difficult-to-measure radionuclides (e.g., pure alpha or beta emitting radionuclides). The reviews

were conducted to verify that the licensee's program assured compliance with

10 CFR 61.55 and 10 CFR 61.56, as required by Appendix G of 10 CFR Part 20. The

inspectors also reviewed the licensee's waste characterization and classification

program to ensure that the waste stream composition data accounted for changing

operational parameters and thus remained valid between the annual sample analysis

updates. This review represented one sample.

b. Findings

No findings of significance were identified..4Shipment Preparation

a. Inspection Scope

The inspectors reviewed shipment packaging, surveying, labeling, marking, placarding, vehicle checks, emergency instructions, disposal manifest, shipping papers provided to

the driver, and licensee verification of shipment readiness for a dry active waste

shipment. The inspectors verified that the receiving licensee was authorized to receive

the shipment packages. The inspectors reviewed the licensee's procedures for loading

and closure. The inspectors observed radiation worker practices to verify that the

workers had adequate skills to accomplish each task and to determine if the shippers

were knowledgeable of the shipping regulations and whether shipping personnel

demonstrated adequate skills to accomplish the package preparation requirements for

public transport with respect to NRC Bulletin 79-19 and 49 CFR Part 172 Subpart H.

The inspectors reviewed the training provided to personnel responsible for the conduct

of radioactive waste processing and radioactive shipment preparation activities. The

review was conducted to verify that the licensee's training program provided training

consistent with NRC and Department of Transportation (DOT) requirements. This

review represented one sample.

b. Findings

No findings of significance were identified..5Shipping Records

a. Inspection Scope

The inspectors reviewed ten non-excepted package shipment manifests completed in years 2005 and 2006 to verify compliance with USNRC and Department of

Transportation requirements (i.e., 10 CFR Parts 20 and 71 and 49 CFR Parts 172

and 173). The inspector reviewed current package preparation or shipping underway

during the inspection. This review represented one sample.

b. Findings

31 No findings of significance were identified..6Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed condition reports, audits, and self-assessments that addressed radioactive waste and radioactive materials shipping program deficiencies since the last

inspection, to verify that the licensee had effectively implemented the corrective action

program and that problems were identified, characterized, prioritized and corrected. The

inspectors also verified that the licensee's self-assessment program was capable of

identifying repetitive deficiencies or significant individual deficiencies in problem

identification and resolution.

The inspectors also reviewed corrective action reports from the radioactive material and shipping programs since the previous inspection, interviewed staff and reviewed

documents to determine if the following activities were being conducted in an effective

and timely manner commensurate with their importance to safety and risk:*Initial problem identification, characterization, and tracking;*Disposition of operability/reportability issues;

Evaluation of safety significance/risk and priority for resolution;

Identification of repetitive problems;

Identification of contributing causes;

Identification and implementation of effective corrective actions;

Resolution of non-cited violations (NCVs) tracked in corrective action system(s);

and*Implementation/consideration of risk significant operational experience feedback.

This review represented one sample.

b. Findings

No findings of significance were identified.2PS3Radiological Environmental Monitoring Program (REMP) and Radioactive Material Control Program (71122.03).1Reviews of Radiological Environmental Monitoring Reports, Data and Quality Control

a. Inspection Scope

The NRC performed a number of confirmatory measurements of water samples to evaluate the licensee's proficiency in collecting and in analyzing water samples for

tritium and other radioactive isotopes. The samples were collected independently by the

inspectors and/or by licensee personnel and sent to the NRC's contract laboratory for

the analysis of tritium. The NRC and licensee obtained these samples from surface

water and groundwater sampling points identified in the licensee's Radiological

Environmental Monitoring Program and from onsite and offsite groundwater monitoring 32 wells. In particular, samples were obtained as part of the licensee's environmental study of tritium and potential groundwater contamination (ADAMS ML062750384) and as part

of an evaluation of leakage from the circulating water blowdown line that is documented

in Section 2PS1 of this report. While tritium was the primary radionuclide of concern, selected samples were also analyzed for gamma emitting radionuclides and for

strontium. The inspectors performed these reviews to assess the licensee's analytical

detection capabilities for radio-analysis of environmental samples and its ability to

accurately quantify radionuclides to an acceptable level of sensitivity. The criteria used

to compare the sample results is provided in Attachment 2, and the results of the

comparisons between the NRC and licensee results is provided in Attachment 3.

The inspectors considered the following activities in evaluating the cause of any comparisons that did not result in an agreement:*re-analysis by licensee or NRC's contract laboratory;*review of licensee's interlaboratory cross check program results; and

review of data for any apparent statistical biases.

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIES4OA1Performance Indicator Verification (71151)Cornerstone: Mitigating Systems and Barrier Integrity

a. Inspection Scope

The inspectors sampled the licensee's submitted materials for performance indicators (PIs) and periods listed below. The inspectors used PI definitions and guidance

contained in Revision 4 of Nuclear Energy Institute Document 99-02, "Regulatory

Assessment Performance Indicator Guideline" to verify the accuracy of the PI data. The

inspectors reviewed selected applicable condition reports and data from logs, licensee

event reports, and work orders for each PI area specified below. The following PIs for

Unit 1 and Unit 2 (4 samples) were reviewed:*Unit 1 Reactor Coolant System Leakage (June 2004 to June 2006)*Unit 2 Reactor Coolant System Leakage (June 2004 to June 2006)

Unit 1 Safety System Functional Failure (October 2004 to June 2006)

Unit 2 Safety System Functional Failure (October 2004 to June 2006)

The documents reviewed during this inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

33 Cornerstones: Occupational and Public Radiation Safety.1Radiation Safety Strategic Area

a. Inspection Scope

The inspectors sampled the licensee's Performance Indicator (PI) submittals for the periods listed below. The inspectors used PI definitions and guidance contained in

Revision 3 of Nuclear Energy Institute Document 99-02, "Regulatory Assessment

Performance Indicator Guideline," to verify the accuracy of the PI data. The following

PIs were reviewed:*Occupational Exposure Control Effectiveness: Units 1 and 2

The inspectors reviewed the licensee's assessment of the PI for occupational radiation safety to determine if indicator related data was adequately assessed and reported

during the previous four quarters. The inspectors compared the licensee's PI data with

the condition report database, reviewed radiological restricted area exit electronic

dosimetry transaction records, and conducted walkdowns of accessible locked high

radiation area entrances to verify the adequacy of controls in place for these areas.

Data collection and analysis methods for PIs were discussed with licensee

representatives to determine if there were any unaccounted for occurrences in the

Occupational Radiation Safety PI, as defined in Revision 3 of Nuclear Energy Institute

Document 99-02, "Regulatory Assessment Performance Indicator Guideline." This

review represented one sample.*Radiological Environmental Technical Specification/Offsite Dose Calculation Manual (RETS/ODCM) Radiological Effluent Occurrences: Units 1 and 2 The inspectors reviewed data associated with the RETS/ODCM PI to determine if the indicator was accurately assessed and reported. This review included the licensee's

condition report database for the previous four quarters to identify any potential

occurrences such as unmonitored, uncontrolled or improperly calculated effluent

releases that may have impacted offsite dose. The inspectors also selectively reviewed

gaseous and liquid effluent release data and the results of associated offsite dose

calculations and quarterly PI verification records generated over the previous four

quarters. Data collection and analyses methods for PIs were discussed with licensee

representatives to determine if the process was implemented consistent with industry

guidance in Revision 3 of Nuclear Energy Institute Document 99-02, "Regulatory

Assessment Performance Indicator Guideline." This review represented one sample.

b. Findings

No findings of significance were identified. However, the inspectors reviewed the adequacy of the licensee's evaluation of abnormal radiological restricted area exit

electronic dosimetry transaction records. Specifically, the records for a condition

identified as "Digi Reset" were reviewed. Based on the licensee's understanding, this "Digi Reset" condition represented an event that indicates the dosimeter was not

functioning for some period of time while the dosimeter was in use. While the 34 dosimeter was not functioning, dose that was received by the worker would not be recorded by the dosimeter. Therefore, this condition could represent

an occurrence in the Occupational Radiation Safety PI as defined in Revision 3 of

Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance

Indicator Guideline." At the time of this inspection, the licensee had not determined the

extent of the issue nor the impact of the condition on the workers dose records. The

licensee planned to perform additional investigations to quantify the duration the

dosimeter was not functioning, the amount of dose that was missed during this time, and

an evaluation of compliance with the requirements specified in Technical Specification 5.7 "Administrative Controls for High Radiation Areas." Therefore, this issue remains

unresolved pending NRC review of the licensee's evaluations, and therefore the issue is

categorized as an Unresolved Item (URI), (05000454/2006004-04;05000455/2006004-

04).4OA2Identification and Resolution of Problems (71152).1Review of Items Entered into the Corrective Action Program:

As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance

issues for follow-up, the inspectors performed screening of all items entered into the

licensee's corrective action program. This was accomplished by reviewing the

description of each new Issue Report and attending selected daily management review

committee meetings. Documents reviewed are listed in the attachment.4OA3Event Follow-Up (71153)Two samples were performed for this inspection module..1Operator Response to Numerous Equipment Failure During Severe Thunderstorm/Lightning Strike

a. Inspection Scope

The inspectors observed and evaluated control room and equipment operator responses to the numerous equipment failure during a severe thunderstorm/lightning strike on July

20, 2006. The inspectors evaluated crew performance in the areas of:*prioritization, interpretation and verification of alarms;*procedure use;

control board manipulations;

supervisor's command and control;

management oversight; and

group dynamics.

Crew performance in these areas was compared to licensee management expectations and procedures. Additional documents reviewed during this inspection are listed in the to this report.

b. Findings

No findings of significance were identified..2Fire Brigade Response to an Auxiliary Building Alarm

a. Inspection Scope

On August 28, 2006, the inspectors responded to the control room and the auxiliary building after hearing a plant announcement of a fire in the auxiliary building. Control

room personnel had received a fire alarm at 11:38 a.m. for the Auxiliary Building

Elevator. The presence of smoke was reported at the 451' level of the auxiliary building

and the fire brigade was dispatched. Offsite fire department assistance was requested

and received.

The fire brigade reported smoke in the area but no fire was observed. The fire brigade found that the elevator brake shaft was very hot so power to the elevator was removed.

Smoke was subsequently cleared and atmospheric testing verified the air was safe to

breath. The event was then terminated and no Emergency Action Level was declared.

The inspectors assessed licensee performance during the event, damage assessment

activities following the event, and the prompt investigation efforts.

b. Findings

No findings of significance were identified.4OA5Other Activities.1(Closed) Unresolved Item (URI) 050000454/2006002-02;050000455/2006002-02:

Licensee Unable to Verify Pipe Integrity Industry experience had shown that the failure of circulating water blow-down line

vacuum breakers resulted in the release of contaminated water into the offsite

environment resulting in groundwater contamination. On February 2, 2006, the licensee

could not demonstrate the integrity of circulating water blowdown line vacuum breakers.

Leakage of water from the blow-down line could result in the release of radioactive

material into the environment via a release path that was undefined in the ODCM and

had no offsite dose estimates.

The licensee conducted additional inspections and analysis of the area surrounding the vacuum breaker vaults. These investigations were evaluated in Section 2PS1.1 of this

report and resulted in a Non-Cited Violation, and the URI is closed..2(Closed) Unresolved Item (URI)05000454/2006002-04

Quantification of Containment Isolation Valve leakage On January 23, 2006, the licensee identified that the Unit 1 Pressurizer liquid sample inboard and outboard containment isolation valves were leaking by.

This condition was

not communicated to the shift manager until two days later. The shift manager 36 subsequently declared both containment isolation valves inoperable and entered the appropriate limiting condition for operations in accordance with Technical

Specifications 3.6.3. Since the condition was discovered two days before, the required

TS action completion time of one hour would have been exceeded. However, a TS

violation exists only if the leakage through the containment isolation valves exceeded 0.6

times the maximum allowable containment leakage rate. The licensee were not able to

quantify the leakage due to existing plant configuration until September, 2006 when Unit

1 shutdown for refueling.

In September 2006, the licensee performed a local leak rate test and determined that the containment isolation valve leakage did not exceed 0.6 times the maximum allowable

containment leakage rate. Therefore, the two containment isolation valves were

operable at the time and no violation of TS existed. This URI is closed..3Pressurized Water Reactor Containment Sump Blockage (TI 2515/166)

a. Inspection Scope

The purpose of this Temporary Instruction was to support Nuclear Regulatory Commission review of licensee's activities in response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized

Water Reactors (PWRs)." This TI required NRC inspectors to verify actions

implemented in response to NRC Generic Letter were complete and where applicable

were programmatically controlled.

The inspectors performed a review in accordance with TI 2515/166 of the licensee's response to GL 2004-02 for Unit 1. The inspectors also reviewed changes to the

licensee's facility and verified they were evaluated in accordance with 10 CFR

Part 50.59. The licensee had received permission to deviate from the schedule in

GL 2004-02 for Unit 1 regarding the downstream effects portion of their modifications.

This portion of the licensee's response to the GL was not modified in the Unit 1

Refueling Outage. As such, TI 2515/166 for Unit 1 remains open.

The inspectors reviewed the licensee's modification packages, attended planning meetings, observed training activities in a recirculation sump mockup, and reviewed

regulatory submittals as part of their preparation activities before the Unit 1 refueling

outage. During the refueling outage the inspectors periodically observed work activities

focusing on the critical attributes selected by the inspectors. For example, the

inspectors compared trash racks, sump screens, and supports to installation drawings.

In addition, the inspectors closely observed Foreign Material Exclusion programs and

practices to ensure FME was not left inside of the new sump screens.b.Evaluation of Inspection Requirements The TI requested the inspectors to include answers to the following questions in this inspection report.1.Did the licensee implement the plant modifications and procedure changes committed to in their GL 2004-02 responses?

37 With the exception of the downstream effects portion of their response the licensee did implement the plant modifications and procedure changes committed to in their GL 2004-02 responses.2.Has the licensee updated its licensing bases to reflect the corrective actions taken in response to GL 2004-02?

The inspectors reviewed the completed 10 CFR Part 50.59 assessments performed by the licensee and verified that the documents contained updates to the UFSAR to be

submitted to the NRC at the next regular update. This is with the exception of the

downstream effects portion of the GL 2004-02 response.

The TI for Unit 1 is not complete. Further inspection is required, specifically, the downstream effects aspects of the ECCS sumps.

c. Findings

No findings of significance were identified.4OA6Meetings.1On October 3, 2006, the resident inspectors presented the inspection results to Mr. D. Hoots and his staff, who acknowledged the findings. The inspectors asked the

licensee whether any materials examined during the inspection should be considered

proprietary. No proprietary information was identified..2Interim Exit Meetings Interim exits were conducted for:

Occupational Radiation Safety Program for radiation monitoring instrumentation and protective equipment and aspects of the effluent monitoring program with Mr.

D. Hoots on July 21, 2006;*Public Radiation Safety Program for radioactive material processing and transportation program and Performance Indicator Verification with Mr. D. Hoots

on August 25, 2006;*Occupational Radiation Safety Program for access control to radiologically significant areas and Al-Low-As-Reasonably-Achievable Planning and Controls (ALARA) programs with Mr. D. Hoots on September 15, 2006;*Occupational Radiation Safety Program Green finding and associated violations of NRC requirements post and control access to High Radiation Areas with

Ms. M. Snow on October 5, 2006; and*Public Radiation Safety with Mr. S. Kerr on October 12, 2006.4OA7Licensee Identified Violations The following violations of very low significance were identified by the licensee and were violations of NRC requirements which met the criteria of Section VI of the NRC

Enforcement Manual, NUREG-1600, for being dispositioned as NCVs.

1 Cornerstone: Initiating Events and Barrier Integrity Byron Station's Operating License Condition 2.C.(6) states, in part, that "The licensee shall implement and maintain in effect all provisions of the approved fire protection

program as described in the SER." Section 9.5.1 of the UFSAR states that "The design

bases, system descriptions, safety evaluat ion, inspection and testing requirements, personnel qualification, and training are described in Reference 1 [the Fire Protection

Report]." Section 3.5.a.5 of the Fire Protection Report states, in part, that access

protected by automatic total flooding gas suppression systems should have electrically supervised self-closing fire doors or should have fire doors that are kept closed and

electrically supervised at a continuously m anned location. Contrary to the above, the licensee failed to have electrically supervised fire door between the diesel generator

rooms and their associated ventilation rooms as the diesel generator rooms were

protected by automatic total flooding carbon dioxide gas suppression systems.

This violation is of very low safety significance because the violation is of low degradation that it only affected suppression, not detection or ignition, and the

suppression system performance and reliabilit y was minimally impacted by the lack of electrically supervised fire door. This issue was entered into the licensee's corrective

action program as IR 513527.

Technical Specification 3.6.7, Spray Additive System, Condition A states that with the spray additive system inoperable, the required action is to restore the system to

operable status within seven days. Contrary to the above, the licensee failed to repair a

pressure boundary leak in the spray additive system within the seven-day allowable

outage time. Specifically, the licensee identified a pressure boundary weld leak in a

ASME Class II pipe of the spray additive system on August 11, 2006. However, they

failed to recognize until September 11, 2006, that the leak rendered the spray additive

system inoperable. The licensee declared t he system inoperable and repaired the leak.

This violation is of very low safety significance because the system does not affect core damage frequently and has no impact on Large Early Release Frequency (LERF). This

issue was entered into the licensee's corrective action program as IR 519173

and 526745.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Hoots, Site Vice President

M. Snow, Plant Manager

B. Adams, Work Control Director

B. Barton, Radiation Engineering Superintendent

L. Doyle, Programs Coordinator

A. Giancatarino, Engineering Director

C. Gregory, RP Instrumentation Coordinator

S. Swanson, Maintenance Director

D. Palmer, Radiation Protection Manager,

W. Grundmann, Regulatory Assurance Manager

W. Kouba, Nuclear Oversight Manager

M. Prospero, Operations Manager

D. Thompson, Technical Support Superintendent

Nuclear Regulatory Commission

R. Skokowski, Chief, Branch 3, Division of Reactor Projects

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

050000454/2006004-04

050000455/2006004-04URIImpact of nonfunctional dosimeters on dose tracking and

Technical Specification compliance (Section 4OA1.1)

Opened and Closed

05000454/2006004-01;

05000455/2006004-01NCVFailure to Maintain Fire Barriers in Accordance with Fire

Protection Program (Section 1R05)

05000454/2006004-02

05000455/2006004-02NCVThe failure to post and control a High Radiation Area (Section 2OS1.1)

05000454/2006004-03

05000455/2006004-03NCVThe failure to evaluate the potential radiological hazard

associated with the leakage of water from the vacuum

breaker valve vault (Section 2PS1.1)

Closed

05000454/2006002-02

05000455/2006002-02URILicensee Unable to Verify Pipe Integrity (Section 4OA5)05000454/2006002-04URIQuantification of Containment Isolation Valve leakage (Section 4OA5)

Discussed

None

LIST OF DOCUMENTS REVIEWED

v t e Inspection Report - Byron - 2006004
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
2006 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 11(0) Security Operator Exams 301(0) Emergency Planning Other