Text

IR 05000454-10-004, 05000455-10-004, 07200068-10-001; 07/01/10 09/30/10; Byron Station, Units 1 & 2; Routine Integrated Inspection Report; Other Activities
	ML103080178
Person / Time
Site:	Byron
Issue date:	11/03/2010
From:	Eric Duncan; Region 3 Branch 3
To:	Pacilio M; Exelon Generation Co, Exelon Nuclear
References
	IR-10-001, IR-10-004
	Download: ML103080178 (53)
	v • d • e

mber 3, 2010

SUBJECT:

BYRON STATION, UNITS 1 AND 2, INTEGRATED INSPECTION REPORT 05000454/2010004; 05000455/2010004; 07200068/2010001

Dear Mr. Pacillo:

On September 30, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Byron Station, Units 1 and 2. The enclosed report documents the results of this inspection, which were discussed on October 8, 2010, with the Mr. Brad Adams, and other members of your staff.

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

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

Based on the results of this inspection, the NRC has determined that one Severity Level IV violation occurred and one NRC-identified finding of very low safety significance was identified.

Both of these involved violations of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as non-cited violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with 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, DC 20555-0001; and the Resident Inspector Office at the Byron Station. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Byron Station. The information that you provide will be considered in accordance with Inspection Manual Chapter 0305.

M. Pacilio -2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Eric R. Duncan, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-454; 50-455;72-068 License Nos. NPF-37; NPF-66

Enclosure:

Inspection Report 05000454/2010004; 05000455/2010004; 07200068/2010001 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 05000454; 05000455; 07200068 License Nos: NPF-37; NPF-66 Report No: 05000454/2010004; 05000455/2010004; 07200068/2010001 Licensee: Exelon Generation Company, LLC Facility: Byron Station, Units 1 and 2 Location: Byron, IL Dates: July 01, 2010 through September 30, 2010 Inspectors: B. Bartlett, Senior Resident Inspector J. Robbins, Resident Inspector R. Ng, Project Engineer M. Learn, Reactor Engineer J. Cassidy, Senior Health Physicist R. Jickling, Senior Emergency Preparedness Inspector C. Thompson, Resident Inspector, Illinois Department of Emergency Management Approved by: E. Duncan, Chief Branch 3 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000454/2010004, 05000455/2010004, 07200068/2010001; 07/01/10 - 09/30/10; Byron

Station, Units 1 & 2; Routine Integrated Inspection Report; Other Activities This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. One SL IV and one green finding were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) of NRC regulations.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Barrier Integrity

Green.

A finding of very low safety significance and an associated Non-Cited Violation (NCV) of Title 10 Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III,

Design Control was identified by the inspectors for the licensees failure to perform adequate evaluations to upgrade the single failure proof crane. Specifically, the inspectors identified six examples where the licensee failed to perform adequate evaluations in accordance with American Society of Mechanical Engineers (ASME)requirements. The licensee documented the conditions in Issue Report (IR) 1099897, and IR 1100062 and initiated actions for calculation revisions and field modifications.

The Fuel Handling Building (FHB) crane was designed to Seismic Category I requirements and the licensee used compliance with ASME NOG-1-2004, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder),

2004, as the design basis for their crane upgrade to a single failure proof crane. The inspectors determined that the failure to perform adequate evaluations was contrary to ASME NOG-1-2004 requirements and was a performance deficiency. The finding was more than minor as it was associated with the Barrier Integrity cornerstone and attribute of design control because a fuel handling building crane heavy load drop can damage the Spent Fuel Pool (SFP) Cooling System or spent fuel cladding. The inspectors evaluated the finding using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and based on a No answer to all of the questions in the Barrier Integrity column of Table 4a, determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources (H.2(c)) because the licensee did not provide adequate oversight of work activities, including contractors, such that design documentation was accurate to support nuclear safety. (Section 4OA5)

Miscellaneous Matters Severity Level IV. A violation of very low safety significance of 10 CFR 72.212 (b)(2)(i)(B), Conditions of a General License Issued under 72.210, was identified by the inspectors for the failure to perform adequate evaluations of the Independent Spent Fuel Storage Installation (ISFSI) pad. Specifically, the inspectors identified four examples where the licensee failed to design the ISFSI pad to adequately support the static and dynamic loads of the stored casks, considering potential amplification of earthquakes through soil-structure interaction (SSI). The licensee documented the conditions in IR 957945 and IR 1102835. As an interim corrective action, the licensee provided a technical paper providing justification for partial loading of the pad with 10 casks.

Because this violation was related to an ISFSI license, it was dispositioned using the traditional enforcement process using Section 6.5.d.1 of the Enforcement Policy. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Dispositioning Screening, because if left uncorrected, a failure of the ISFSI pad could lead to a more significant safety concern. This violation is being treated as a NCV consistent with Section 2.3.2 of the NRC Enforcement Policy. The violation screened as having very low safety significance (Severity Level IV). (Section 4OA5)

Licensee-Identified Violations

Violations of very low safety significance that were identified by the licensee have been reviewed by inspectors. Corrective actions planned or taken by the licensee have been entered into the licensees corrective action program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at or near full power during this inspection period with one exception. On September 22, 2010, the unit was operating at full power with one main feedwater pump out-of-service for maintenance when the A main feedwater pump tripped. Power was reduced to approximately 50 percent to maintain steam generator water levels. Following repairs to the main feedwater pump oil system the pump was restarted and the unit returned to full power the following day.

Unit 2 operated at or near full power during this inspection period with no major problems.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

Main article: IP 71111.04

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

Unit 2 Train B Residual Heat Removal (RH) while the Unit 2 Train B RH was out of service for scheduled maintenance;

Portions of both Unit 1 Auxiliary Feedwater (AF) trains within the normally inaccessible AF Pipe Tunnel;

Unit 2 Train B Containment Spray during emergent work on valve 2SI8811A; and

Essential Service Water Make-Up Pumps and Discharge Lines during licensee surveillance.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, the Updated Final Safety Analysis Report (UFSAR), Technical Specification (TS) requirements, outstanding work orders (WOs), issue reports (IRs),and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program (CAP) with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

These activities constituted four partial system walkdown samples as defined in IP 71111.04-05.

b. Findings

No findings of significance were identified.

.2 Semi-Annual Complete System Walkdown

a. Inspection Scope

On August 10, 2010, the inspectors performed a complete system alignment inspection of the Unit 1 Component Cooling while the Unit 0 Component Cooling was aligned, to verify the functional capability of the system. This system was selected because it was considered both safety significant and risk significant in the licensees probabilistic risk assessment. The inspectors walked down the system to review mechanical and electrical equipment line ups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of a sample of past and outstanding WOs was performed to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the CAP database to ensure that system equipment alignment problems were being identified and appropriately resolved.

Documents reviewed are listed in the Attachment to this report.

These activities constituted one complete system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Routine Resident Inspector Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

Fire Zone 9.2-1 & 9.3-1, 1A Diesel Generator & Day Tank Room;

Fire Zone 11.5, Auxiliary Building 401' Elevation, General Area; and

Fire Zone 11.3, Auxiliary Building 364' Elevation, General Area.

The inspectors reviewed these areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the Attachment, 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 fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP. Documents reviewed are listed in the Attachment to this report.

These activities constituted three quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings of significance were identified.

1R07 Annual Heat Sink Performance

Main article: IP 71111.07

.1 Heat Sink Performance

a. Inspection Scope

The inspectors reviewed the licensees testing of Unit 1 Train B Safety Injection Pump Oil Cooler and Room Coolers heat exchangers to verify that potential deficiencies did not mask the licensees ability to detect degraded performance, to identify any common cause issues that had the potential to increase risk, and to ensure that the licensee was adequately addressing problems that could result in initiating events that would cause an increase in risk. The inspectors reviewed the licensees observations as compared against acceptance criteria, the correlation of scheduled testing and the frequency of testing, and the impact of instrument inaccuracies on test results. The inspectors also verified that test acceptance criteria considered differences between test conditions, design conditions, and testing conditions. Documents reviewed for this inspection are listed in the Attachment to this report.

This annual heat sink performance inspection constituted one sample as defined in IP 71111.07-05.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

Main article: IP 71111.11

.1 Resident Inspector Quarterly Review

Main article: IP 71111.11Q

a. Inspection Scope

On July 27, 2010, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification examinations to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

licensed operator performance;

crews clarity and formality of communications;

ability to take timely actions in the conservative direction;

prioritization, interpretation, and verification of annunciator alarms;

correct use and implementation of abnormal and emergency procedures;

control board manipulations;

oversight and direction from supervisors; and

ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator requalification program sample as defined in IP 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

.1 Routine Quarterly Evaluations

Main article: IP 71111.12Q

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk-significant systems:

Pressurizer Breaker Tripped;

Unit 1 Essential Service Water (SX) Valve Leakby; and

Unit 1 Component Cooling Water System Cross Unit Leakage.

The inspectors reviewed events, such as ineffective equipment maintenance that resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

implementing appropriate work practices;

identifying and addressing common cause failures;

scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

characterizing system reliability issues for performance;

crediting unavailability for performance;

trending key parameters for condition monitoring;

ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted three quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

Unit 1 Component Cooling Water System Cross Tie Valve Leakby;

Trip of Unit 1A Main Feedwater Pump and down power to 50 percent;

Pressurizer Code Safety Valve 2A Seat Leakage; and

Unit 1 and Unit 2 Solid State Protection System Emergent Testing Due to Operating Experience from another Unit.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Documents reviewed are listed in the attachment to this report.

These maintenance risk assessments and emergent work control activities constituted four samples as defined in IP 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

Main article: IP 71111.15

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

Unit 1 Train B Diesel Generator Fuel Oil Storage Tank Level Measurement Uncertainty Issues;

Potential Design Vulnerability for Auxiliary Feedwater;

Unit 1 Train B Safety Related Relay K611 Contacts 13-14 High Resistance; and

Unit 2 Train A RH Room Cooler Fan Motor Bearing to Shaft Clearance May be Beyond Standard Clearance.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and Updated Final Safety Analysis Report (UFSAR) to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations.

Documents reviewed are listed in the Attachment to this report.

This operability inspection constituted four samples as defined in IP 71111.15-05.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

Main article: IP 71111.18

.1 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the following temporary modification:

Temporary Pump in Unit 1 SX Room Sump The inspectors compared the temporary configuration changes and associated 10 CFR 50.59 screening and evaluation information against the design basis, the UFSAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected system. The inspectors also compared the licensees information to operating experience information to ensure that lessons learned from other utilities had been incorporated into the licensees decision to implement the temporary modification. The inspectors, as applicable, performed field verifications to ensure that the modification was installed as directed; the modification operated as expected; modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modification did not impact the operability of any interfacing systems. Lastly, the inspectors discussed the temporary modification with operations, engineering, and training personnel to ensure that the individuals were aware of how extended operation with the temporary modification in place could impact overall plant performance. Documents reviewed in the course of this inspection are listed in the Attachment to this report.

This inspection constituted one temporary modification sample as defined in IP 71111.18-05.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance (PM) testing activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

Unit 0 Train B SX Makeup Pump Battery Charge Control Circuit Card Replacement;

Unit 1 Diesel Generator Fuel Oil Level Switch Replacement;

Unit 2 Train A Emergency Core Cooling System Sump Suction Isolation Valve 2SI8811A; and

AF Valve 2AF0006A Leaking By and Will Not Move.

These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TS, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment to this report.

This inspection constituted four post-maintenance testing samples as defined in IP 71111.19-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

Main article: IP 71111.22

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

0BOSR IS-SA-1 Security DG Semi-Annual Test (Routine);

Set up a 480V MCC Bucket for use in Cubicle B4 of Safety-Related MCC 231X3 (Routine);

Stroke Time Test of Unit 1 SX System Valves (In-Service Test);

Monthly Test of Unit 2 Reactor Containment Fan Coolers (Routine);

Stroke Time Test of Chilled Water Isolation Valves (In-Service Test); and

Monthly Operability Test of 0A SX Make Up Pump (Routine).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

did preconditioning occur;

were the effects of the testing adequately addressed by control room personnel or engineers prior to the commencement of the testing;

were acceptance criteria clearly stated, demonstrated operational readiness, and consistent with the system design basis;

was plant equipment calibration correct, accurate, and properly documented;

were as-left setpoints within required ranges; and the calibration frequencies in accordance with TSs, the UFSAR, procedures, and applicable commitments;

was measuring and test equipment calibration current;

was test equipment within the required range and accuracy; and were applicable prerequisites described in the test procedures satisfied;

did test frequencies meet TS requirements to demonstrate operability and reliability; were tests performed in accordance with the test procedures and other applicable procedures; were jumpers and lifted leads controlled and restored where used;

were test data and results accurate, complete, within limits, and valid;

was test equipment removed after testing;

was testing, where applicable for inservice testing activities, performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers code, and were reference values consistent with the system design basis;

were test results not meeting acceptance criteria addressed with an adequate operability evaluation or the system or component was declared inoperable;

were reference setting data, where applicable for safety-related instrument control surveillance tests, accurately incorporated in the test procedure;

were actual conditions encountering high resistance electrical contacts such that the intended safety function could still be accomplished;

did prior procedure changes provide an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

was equipment returned to a position or status required to support the performance of its safety functions; and

were all problems identified during the testing appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted four routine surveillance testing samples and two inservice testing samples as defined in IP 71111.22, Sections -02 and -05.

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP2 Alert and Notification System Evaluation

Main article: IP 71114.02

.1 Alert and Notification System Evaluation

a. Inspection Scope

The inspectors reviewed documents and conducted discussions with Emergency Preparedness (EP) staff and management regarding the operation, maintenance, and periodic testing of the Alert and Notification System (ANS) in the Byron Station's plume pathway Emergency Planning Zone. The inspectors reviewed monthly trend reports and the daily and monthly operability records from May 2008 through June 2010.

Information gathered during document reviews and interviews was used to determine whether the ANS equipment was maintained and tested in accordance with Emergency Plan commitments and procedures. Documents reviewed are listed in the Attachment to this report.

This alert and notification system inspection constituted one sample as defined in IP 71114.02-05.

b. Findings

No findings of significance were identified.

1EP3 Emergency Response Organization Augmentation Testing

Main article: IP 71114.03

.1 Emergency Response Organization Augmentation Testing

a. Inspection Scope

The inspectors reviewed and discussed with plant Emergency Preparedness (EP)management and staff the emergency plan commitments and procedures that addressed the primary and alternate methods of initiating an Emergency Response Organization (ERO) activation to augment the on shift ERO as well as the provisions for maintaining the stations ERO qualification and team lists. The inspectors reviewed reports and a sample of corrective action program records of unannounced off-hour augmentation tests and pager tests, which were conducted between December 2008 and May 2010, to determine the adequacy of the drill critiques and associated corrective actions. The inspectors also reviewed a sample of the EP training records of approximately 24 ERO personnel, who were assigned to key and support positions, to determine the status of their training as it related to their assigned ERO positions.

Documents reviewed are listed in the Attachment to this report.

This emergency response organization augmentation testing inspection constituted one sample as defined in IP 71114.03-05.

b. Findings

No findings of significance were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies

Main article: IP 71114.05

.1 Correction of Emergency Preparedness Weaknesses and Deficiencies

a. Inspection Scope

The inspectors reviewed a sample of Nuclear Oversight (NOS) staffs 2009 and 2010 audits of the Byron Station's emergency preparedness program to determine whether the independent assessments met the requirements of 10 CFR 50.54(t). The inspectors also reviewed samples of corrective action program records associated with the 2009 biennial exercise, as well as various EP drills conducted in 2009 and 2010, in order to determine whether the licensee fulfilled drill commitments and to evaluate the licensees efforts to identify and resolve identified issues. The inspectors reviewed a sample of EP items and corrective actions related to the facilitys EP program and activities to determine whether corrective actions were completed in accordance with the sites corrective action program. Documents reviewed are listed in the Attachment to this report.

This correction of emergency preparedness weaknesses and deficiencies inspection constituted one sample as defined in IP 71114.05-05.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

Main article: IP 71114.06

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on September 22, 2010, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the Simulator/Control Room and Technical Support Center to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures.

The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the corrective action program. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the to this report.

This emergency preparedness drill inspection constituted one sample as defined in IP 71114.06-05.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstones: Occupational and Public Radiation Safety

2RS5 Radiation Monitoring Instrumentation

Main article: IP 71124.05

This inspection constituted a partial sample as defined in IP 71124.05-5.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed the plant UFSAR to identify radiation instruments associated with monitoring area radiological conditions including airborne radioactivity, process streams, effluents, materials/articles, and workers. Additionally, the inspectors reviewed the instrumentation and the associated Technical Specification requirements for post-accident monitoring instrumentation including instruments used for remote emergency assessment.

The inspectors reviewed a listing of in-service survey instrumentation including air samplers and small article monitors (SAMs), along with instruments used to detect and analyze workers external contamination. The inspectors also reviewed personnel contamination monitors (PCMs) and a portal monitor including whole-body counters (WBCs) to detect workers internal contamination. From a review of the list, the inspectors evaluated whether the licensee had an adequate number and type of instruments to support plant operations.

The inspectors reviewed selected independent evaluation reports of the radiation monitoring program since the last inspection. This review included audits of the licensees onsite calibration facility.

b. Findings

No findings of significance were identified.

.2 Walkdowns and Observations (02.02)

a. Inspection Scope

The inspectors performed walked downs of following effluent radiation monitoring systems (liquid and airborne system):

High Range Containment Monitor;

Liquid Effluent Radiation Monitors;

Gaseous Effluent Radiation Monitors; and

Auxiliary Building Vent Stack Effluent Monitors.

The inspectors also evaluated the flow measurement devices and all accessible point-of-discharge liquid and gaseous effluent monitors of the selected systems.

The inspectors also assessed whether effluent/process monitor configurations aligned with ODCM descriptions and whether any monitors were out of service or degraded.

The inspectors selected several portable survey instruments during the inspection to verify calibration and source check stickers and to assess instrument material condition and operability. The inspectors observed licensee staff demonstrate source checks for various types of portable survey instruments and determined that the source checks were performed at the appropriate scales (low and high range). The inspectors walked down multiple ARMs and continuous air monitors (CAMs) to determine whether they were appropriately positioned to the areas they were intended to monitor. Selectively, the inspectors compared monitor response via local or remote control room indications with actual area conditions for consistency.

The inspectors selected several PCMs, portal monitors, and SAMs and evaluated whether the periodic source checks were performed in accordance with the manufacturers recommendations and licensee procedures.

b. Findings

No findings of significance were identified.

.3 Calibration and Testing Program (02.03)

Process and Effluent Monitors

a. Inspection Scope

The inspectors selected several effluent monitor instruments (such as gaseous and liquid) and assessed whether channel calibration and functional tests were performed consistent with radiological effluent Technical Specifications (RETS)/ODCM. The inspectors evaluated whether:

(a) the licensee calibrated its monitors with National Institute of Standards and Technology (NIST) traceable sources;

(b) the primary calibrations adequately represented the plant nuclide mix;

(c) when secondary calibration sources were used, the sources were verified by the primary calibration; and

(d) the licensees channel calibrations encompassed the instruments alarm setpoints.

The inspectors reviewed calibration documentation for portable survey instruments and Area Radiation Monitors (ARMs) that included review of the detector measurement geometry and calibration methods. Additionally, inspectors observed the demonstration of the instrument calibrator.

Main article: IP 71151

.1 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled licensee submittals for the Unit 1 and Unit 2 Safety System Functional Failures performance indicator (PI) for the period from April 1, 2009 through June 1, 2010. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, dated October 2009, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73", were used. The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance WOs, IRs, event reports and NRC Integrated Inspection Reports for the period of April 1, 2009 through June 1, 2010 to validate the accuracy of the submittals. The inspectors also reviewed the licensees IR database to determine if any problems had been identified with the PI data collected or transmitted for this indicator. Documents reviewed are listed in the to this report.

This inspection constituted two safety system functional failures samples as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

.2 Reactor Coolant System Leakage

a. Inspection Scope

The inspectors sampled licensee submittals for the Unit 1 and Unit 2 RCS Leakage PI for the period from July 1, 2009 through June 1, 2010. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, dated October 2009, were used. The inspectors reviewed the licensees operator logs, RCS leakage tracking data, IRs, event reports and NRC Integrated Inspection Reports for the period of July 1, 2009 through June 1, 2010, to validate the accuracy of the submittals.

Main article: IP 71152

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

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: the complete and accurate identification of the problem; that timeliness was commensurate with the safety significance; that evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrence reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the attached List of Documents Reviewed.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily IR packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow-Up Inspection: Diesel Oil Storage Tank Level Indication

a. Inspection Scope

During a review of items entered in the licensees CAP, the inspectors recognized a corrective action item documenting an unplanned entry into Limiting Condition for Operation Action Requirement; 3.8.3 Condition A. This Limiting Condition for Operation (LCO) is associated with stored diesel fuel requirements for on-site emergency diesel generators.

Following calibration of level instrument gauges, site staff identified that current tank levels were below their administrative limits. The tanks were promptly filled to a point above their administrative limits. Analyses were requested by the Operations group to support a past operability/reportability determination. Engineering Change (EC) 381086 was initiated to address this issue. The EC concluded that the actual fuel level in the tanks exceeded the Technical Specification required value of 44,000 gallons and was sufficient for 7 days of operation. Based on this evaluation the licensee determined that the fuel tanks were operable and the event was not reportable.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings of significance were identified.

4OA3 Follow-up of Events and Notices of Enforcement Discretion

Main article: IP 71153

.1 (Closed) Licensee Event Report 0500455/2009-001-01: Late Entry Into TS Condition

Associated with Reactor Coolant System Leakage Characterization Resulting in a Condition Prohibited by TSS On June 24, 2009, during a routine containment entry at power licensee personnel identified a pinhole leak (one drop every 5 minutes) on a welded connection inside of Unit 1 containment (IR 934800). The welded connection was on line 2PS01BB and the line was 3/8-inch in diameter. This line is a pressurizer liquid sample line and is a non safety-related non-ASME code, Class D pipe. The licensee verified that upstream valve 2PS9350B (between the leak and the RCS pressurizer) was closed and that both downstream containment isolation valves were closed. Based on the upstream valve being closed and in the Shift Managers opinion being isolated, and with the leakage being not significant, the leak was not considered by licensee personnel to be RCS pressure boundary leakage.

As discussed in NRC Inspection Report 05000454/4552009-004, the inspectors reviewed this Licensee Event Report (LER) and issued NCV 05000454/2009003-01.

Subsequently, the licensee had submitted a response which disagreed with the NRC position that this was pressure boundary leakage. NRC personnel followed the applicable procedures for a disputed violation. After discussions with senior licensee management, the licensee stated that they did not disagree with the violation.

The inspectors reviewed the LER and concluded it was completed in accordance with 10 CFR 50.73. Therefore, this LER is closed.

This event follow-up review constituted one sample as defined in IP 71153-05.

4OA5 Other Activities

.1 (Closed) Temporary Instruction 2515/179, Verification of Licensee Responses to NRC

Requirement for Inventories of Materials Tracked in the National Source Tracking System Pursuant to Title 10, Code of Federal Regulations, Part 20.2207 (10 CFR 20.2207)

a. Inspection Scope

The inspectors confirmed that the licensee has reported the initial inventories of sealed sources pursuant to 10 CFR 20.2207 and verified that the National Source Tracking System database correctly reflected the Category 1 and 2 sealed sources in custody of the licensee. Inspectors interviewed personnel and performed the following:

Reviewed the licensees source inventory;

Verified the presence of any Category 1 or 2 sources;

Reviewed procedures for and evaluated the effectiveness of storage and handling of sources;

Reviewed documents involving transactions of sources; and

Reviewed the adequacy of licensee maintenance, posting, and labeling of nationally tracked sources.

b. Findings

No findings of significance were identified.

2. Preoperational Testing of an Independent Spent Fuel Storage Facility Installation at

Operating Plants (60854.1)

b. Inspection Scope

(1) Control of Heavy Loads The inspectors initiated a review of the licensees crane and heavy loads program with regards to Independent Spent Fuel Storage Facility Installation (ISFSI) operations in 2009 documented in Byron Station Inspection Report Nos. 05000454/2009004; 05000455/2009004. The inspectors opened two Unresolved Items (URIs):

URI 005000454/2009004-04;05000455/2009004-04 Use of Friction in Design of Fuel Handling Building (FHB) Crane to Single Failure Proof and URI 05000454/2009004-05; 05000455/2009004-05 Unresolved Technical Concerns on Design of Seismic Restraint on FHB Crane Trolley. These URIs were opened by the inspectors to follow up and complete a review of the design documents related to the FHB crane upgrade to single failure proof. Specifically, not all of the licensees design and modification documents required to complete the inspection were complete at the conclusion of the IR 05000454/2009004; 05000455/2009004 inspection period.

As part of the modifications in preparations to load fuel in the ISFSI, the licensee upgraded the 125 ton capacity overhead crane in the fuel handling building to a single failure proof crane in compliance with the NRC guidance, NUREG 0612, NUREG 0554 and ASME-NOG-1-2004.

During this inspection period, the inspectors have completed their review of documentation associated with the FHB crane. The review included structural evaluations associated with the seismic design of the new trolley, hoist/reeving equipment, miscellaneous components, crane bridge girders, supporting structural steel, modifications affecting the operating plant, floor loading in the spent fuel pool (SFP) and other cask storage areas, and the cask and canister stack up configurations. The associated safety evaluations and screenings were also reviewed.

(2) Dry Run Activities During this inspection period, the licensee performed the balance of preoperational dry run activities in order to fulfill the Certificate of Compliance (CoC). The NRCs inspection of prior preoperational dry run activities were performed during the first and second quarter of 2009, and were documented in Byron Station Inspection Report Nos. 05000454/2009004; 05000455/2009004. NRC inspectors were onsite to observe dry run activities July 28 through July 30, 2010, August 11 through August 12, 2010, and August 20, 2010. The inspection consisted of observations of dry run activities the licensee performed utilizing the Holtec HI-STORM 100 storage system including multi-purpose canister (MPC) wet operations, transfer of the MPC from the transfer cask (HI-TRAC) to the storage cask (HI-STORM), and document review. The inspectors reviewed loading and unloading procedures not previously reviewed as documented in Inspection Report Nos. 05000454/2009004; 05000455/2009004 to ensure that they contained commitments and requirements specified in the CoC license, the CoC TS, the Holtec Final Safety Analysis Report (FSAR), and 10 CFR Part 72. The inspectors also observed the licensees pre-job briefings. The inspectors determined whether the licensee conducted these meetings in a professional manner and the necessary items to enhance safety were discussed. The inspectors verified that Radiation Protection staff attended pre-job briefs and provided insights into working conditions and as low as reasonably achievable (ALARA) practices.

The inspectors observed the loading and unloading of dummy fuel bundles into the MPC basket. The licensee demonstrated removal of a dummy fuel assembly from the SFP storage rack, placement of the assembly into the MPC, and retrieval of the fuel assembly from the MPC to the SFP rack. The inspectors observed transfer of the MPC from the HI-TRAC cask to the HI-STORM in a restrained support structure in the FHB and the subsequent movement of the HI-STORM outside of the FHB Building. The inspectors verified that proper controls were in place during the transfer of the canister. The inspectors verified adequate communication and team work between departments and adherence to procedures.

(3) Fuel Selection The inspectors reviewed the licensees program associated with fuel characterization and selection for storage. The inspectors reviewed cask fuel selection packages to verify that the licensee was loading fuel in accordance with the CoC TS. The licensee did not plan to load any damaged fuel assemblies during this initial campaign.

(4) Radiation Protection The inspectors evaluated the licensees RP Program pertaining to the operation of the ISFSI. The inspectors observed licensee RP technicians simulate dry run activities and interviewed technicians to verify their knowledge regarding the scope of the work and the radiological hazards associated with transfer and storage of spent fuel.

(5) Training The inspectors reviewed the licensees training program, which consisted of classroom and on-the-job training to ensure involved staff was adequately trained for the job they were responsible to perform. The inspectors also reviewed training records and qualifications of individuals performing work activities associated with the ISFSI. The inspectors interviewed licensee personnel to verify that they were knowledgeable in the scope of work that was being performed.

c. Findings

(1) Failure to Perform Adequate Evaluation for Crane Upgrade Introduction A finding of very low safety significance (Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control was identified by the inspectors for failure to perform adequate evaluations to upgrade their single failure proof crane. Specifically, for evaluations of the fuel handling building crane and crane support structure, the licensee failed to comply with American Society of Mechanical Engineers (ASME) NOG-1-2004, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), 2004, as previously committed to. The licensee documented the conditions in IR Nos. 1099897 and 1100062 and initiated actions for calculation revisions and field modifications.

Description The FHB crane was designed to Seismic Category I requirements. The licensee used compliance with NOG-1-2004, as the basis for its upgrade to single failure proof. ASME NOG-1-2004 was endorsed by the NRC per Regulatory Information Summary (RIS)2005-25, Supplement 1, as an acceptable method for satisfying the guidelines of NUREG-0554 for single failure proof cranes. Inspectors reviewed licensee documents associated with URIs 05000454/2009-004-05; 05000455/2009-004-05 Unresolved Technical Concerns on Design of Seismic Restraint on FHB Crane Trolley.

During a review of calculations for the crane and crane support structure, the inspectors identified six examples where the licensee failed to meet the requirements in 10 CFR Part 50, Appendix B, Criterion III, Design Control.

1. Calculation 36272-12, Revision 15 (10/19/09), Single Failure Proof Trolley Seismic

Analysis:

The calculation listed partial weights for the trolley. The inspectors identified that when added together, the total weight resulted in only 59,000 pounds compared to the actual trolley weight of 68,000 pounds based on certifications attached with calculation 36272-35, Revision 3. This error was non-conservative.

The licensee documented the discrepancy in IR No. 1099897.

2. Calculation 36272-12, Revision 15 (10/19/09), Single Failure Proof Trolley Seismic

Analysis:

The inspectors identified that the analysis results were significantly different than those in the previous revision (Revision 9, 6/1/09). Upon questioning, the inspectors learned that during the revision the licensee had corrected the boundary conditions in the trolley/bridge analytical model to conform to the ASME NOG-1-2004, Section 4153.6 requirements. This discrepancy was identified during the revision in response to the NRC questions. The licensee documented the discrepancy in IR No. 1099897.

3. Calculation 36272-12, Revision 15 (10/19/09), Single Failure Proof Trolley Seismic

Analysis:

ASME NOG-1-2004, Section 4153.7 requires that two separate loading conditions including credible critical load on hook and no load on hook shall be analyzed. It also requires that two positions of the loaded hook including hook up and hook down be analyzed. The licensee trolley analysis did not address the no load on hook condition or the loaded hook down position. The licensee documented the discrepancy in IR No. 1099897.

4. The inspectors identified a number of errors in calculation 36272-02, Revision 11

(10/19/09), Exelon Byron and Braidwood Hoist/Reeving Equipment Calculation: In Section 5.3.7 of the calculation, the subsections a, b, and c, calculated stresses for upper block torque arm parts identified as #9 and #10 were incorrect because of the incorrect definition of variables. The licensee used Mathcad software for computations. The inspectors identified that the variables used to define properties such as the thickness, edge distance, and effective widths for parts 9 and 10 were transposed leading to incorrect computations. Also, in subsection b, a multiplier of 2 was incorrectly applied to a term in the equation denominator resulting in a lower calculated stress. Similar errors were also identified in Section 6.2.7 of the calculation. The licensee documented the discrepancies in IR No. 1099897. The revised calculation (Revision 16, 7/9/10) indicated that the stresses increased in some cases but were found to be within the allowable limits.

5. Calculation 36272-12, Revision 15 (10/19/09), Single Failure Proof Trolley Seismic

Analysis:

ASME NOG-1-2004, Section 4153.5 provides criteria to be met in order to represent the crane as a separate model, decoupled from the crane runway. The inspectors identified that the trolley analysis was based on a bridge / trolley model decoupled from the crane runway but did not demonstrate that the ASME NOG-1-2004 criteria were met. The licensee documented the discrepancy in IR 1099897. The licensee subsequently performed new analyses using a coupled crane/support structure model.

6. As part of the crane upgrade, the licensee installed a new trolley evaluated in

calculation 36272-12. However, the licensee did not evaluate the crane bridge girders and the crane support structure for the loads resulting from the new trolley analysis. In response to the NRC questions, the licensee determined that their existing bridge and support structure calculations were not adequate and decided to perform new calculations. The licensee documented the discrepancy in IR No. 1100062. The licensee subsequently performed new analyses for the crane and the support structure. Based on results of the new analyses, modifications were required to the crane runway rail clips. The results also required significant revisions to the trolley seismic calculations and modification to the bolts supporting the auxiliary hoist frame.

Upon identification by the inspectors, the licensee documented the above examples in IRs 1099897, and 1100062. The crane was not operational as an upgraded single failure proof crane during this period. Resolution of the above items resulted in the licensee performing a number of new calculations and issuing major revisions to the existing calculations demonstrating adequacy of the design after installation of the identified modifications. The crane was converted to single failure proof following additional calculations and modifications.

Analysis The inspectors determined that the licensees failure to perform adequate evaluations to upgrade their single failure proof crane was contrary to the design control measures per 10 CFR Part 50 Appendix B requirements and was a performance deficiency.

The finding was determined to be more than minor because the finding was associated with the Barrier Integrity cornerstone attribute of Design Control and affected the cornerstone objective of providing reasonable assurance that a physical design barrier (fuel cladding) protects the public from radionuclide releases caused by accidents or events. Specifically, the failure to perform adequate evaluations of the fuel handling building crane and crane support structure affected the licensees ability to provide reasonable assurance that the fuel cladding would be maintained if a heavy load was dropped in the FHB which could cause damage to spent fuel.

The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b and 4a for the Barrier Integrity Cornerstone. The finding was more than minor as it was associated with the Barrier Integrity cornerstone and attribute of design control because a fuel handling building crane heavy load drop can damage the Spent Fuel Pool (SFP) Cooling System or spent fuel cladding. Since the finding was a design qualification deficiency confirmed not to result in a heavy load drop and no damage to spent fuel, it was screened as a Green finding. Therefore, the finding was determined to be of very low safety significance (Green).

The inspectors identified a Human Performance, Resources (H.2.c) cross-cutting aspect associated with this finding. The licensee did not provide adequate personnel and other resources necessary for providing complete and accurate design documentation in order to assure nuclear safety. Specifically, the licensee failed to provide complete and accurate design calculations and documentation for establishing structural adequacy of the crane components and the crane support structure for the crane upgrade to single failure proof.

Enforcement Title 10 CFR 50, Appendix B, Criterion III, Design Control states, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above,

1. On October 19, 2009, in calculation 36272-12, Revision 15, the inspectors

determined that the licensees design control measures failed to verify accuracy of the trolley weight used. Specifically, the trolley weight used was 59,000 pounds while the actual weight based on weighing certification was 68,000 pounds.

2. On June 6, 2009, in calculation 36272-12, Revision 9, the inspectors determined that

the licensees design control measures failed to verify accuracy of the boundary conditions used for the trolley/bridge analytical model. Specifically, the boundary conditions used in the analytical model for the crane trolley / bridge evaluation did not conform to the requirements per Section 4153.6 of ASME NOG-1-2004.

3. On October 19, 2009, in calculation 36272-12, Revision 15 the inspectors

determined that the licensees design control measures failed to verify that all the applicable crane loading conditions were evaluated. Specifically, ASME NOG-1-2004, Section 4153.7 requires that two separate loading conditions including credible critical load on hook and no load on hook shall be analyzed. It also requires that two positions of the loaded hook including hook up and hook down be analyzed. The licensee calculation did not evaluate the loading conditions involving no load on hook nor the loaded hook down position.

4. On October 19, 2009, in calculation 36272-02, Revision 11, the inspectors

determined that the licensees design control measures failed to verify the accuracy of the calculations used to demonstrate structural adequacy of trolley components.

Specifically, errors were found in Section 5.3.7 of the calculation, subsections a, b, and c, where the calculated stresses for parts identified as #9 and #10 were incorrect because of the incorrect definition of variables such as the thickness, the edge distance, and effective width. Also, in subsection b, a multiplier of 2 was incorrectly applied to a term in the equation denominator resulting in a lower calculated stress.

Similar errors were also identified in Section 6.2.7 of the calculation.

5. On October 19, 2009, in calculation 36272-12, Revision 15 the inspectors

determined that the licensees design control measures failed to verify that use of a crane model decoupled from the crane runway and support structure was justified.

Specifically, while using a crane model decoupled from the runway, the licensee evaluation did not demonstrate that the decoupling criteria for the crane runway per ASME NOG-1-2004, Section 4153 were satisfied.

6. On October 19, 2009, in calculation 36272-12, Revision 15, the inspectors

determined that the licensees design control measures failed to verify adequacy of the crane bridge girders and the support structure. Specifically, the bridge girders and FHB crane support structure were not evaluated for loads resulting from the new analysis based on the crane upgrade.

Because this matter was of very low safety-significance (Green), and has been entered into the licensees corrective action program under IR Nos. 1099897, and 1100062; this violation is being treated as a NCV consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2010004-01; 05000455/2010004-01).

This NRC identified violation closes URI 05000454/2009004-05; URI 05000455/2009004-05 Unresolved Technical Concerns on Design of Seismic Restraint on FHB Crane Trolley.

.3 Methodology Used in the HI-STORM/HI-TRAC Stack-up Evaluation May Be

Inadequate

Introduction:

A URI was identified by the inspectors regarding the methodology used to transfer the MPC from the HI-TRAC to the HI-STORM. Specifically, the inspectors identified a number of concerns regarding the overall methodology of the calculation.

Description:

Calculation 2.4.3-BYR08-070, Revision 2 (7/20/09) evaluated adequacy of the transfer configuration used for vertical transfer of the MPC from the HI-TRAC to the HI-STORM.

The transfer includes the HI-TRAC placed on top of the HI-STORM with a mating device interposed between the two. All three components are placed on top of a trolley (low profile transporter) that can move along rails on the floor of the FHB. The height of the two casks placed on one another is greater than 30 feet.

A seismic analysis of the configuration was performed by the licensees contractor, Holtec International using time history seismic input and Visual Nastran computer code.

The analysis model used involved an assemblage of multiple freestanding bodies responding to the input seismic motion with friction at various contact surfaces acting as the resisting forces.

The inspectors identified a number of concerns regarding the overall methodology of relying on friction as a restraining force: the lack of benchmarking of the computer code, no apparent prior evaluation by NRC, the method used for application of seismic input, the coefficient of friction values used, and the assumptions regarding coefficient of restitution and damping, etc. The concerns were discussed with the Division of Spent Fuel Storage and Transportation (DSFST) staff and the licensee. In response, the licensee decided to abandon the plan to use freestanding stack-up configuration and instead provided physical restraints for their current ISFSI campaign. The inspectors did not make a determination on the findings or significance of these concerns by the end of the inspection. The inspectors also identified that other licensees were planning to use a similar stack-up configuration. Due to the complexity of the analysis and the potential safety consequences, a Region III Technical Assistance Request (TAR) was issued to the DSFST Headquarters office. The TAR requested staffs review of a similar analysis performed by another licensee and also requested additional guidance for the inspectors. The licensee documented the inspectors concerns in IR Nos. 1031363 and 1049402.

This issue will be a URI pending further review of the calculation by the inspectors after DSFST completes its review of the TAR and provides the necessary inspection guidance (URI 05000454/2010004-02; 05000455/2010004-02).

(2) Review of 10 CFR 72.212(b) Evaluations at Operating Plants

a. Inspection Scope

The inspectors reviewed the licensees ISFSI pad evaluations for compliance with requirements discussed in 10 CFR 72.212 (b)(2)(i)(B) during ISFSI inspections in 2009.

The inspectors identified that evaluations in calculation 2.4.3-BYR08-27 did not include the effect of partial/sequential loading on concrete design and the soil bearing pressure.

URI 07200068/2009001-01 Pad Structural Evaluation Revisions to Address Partial Sequential Loading was opened to track the issue.

During review of ISFSI pad calculations at another operating plant, inspectors identified an issue of concern regarding the licensees evaluation of the ISFSI pad. A similar calculation was used at the Byron Station and the licensee entered the issue into its corrective action program. Unresolved Item 07200068/2009001-02 Use of Methodology Described in NUREG/CR6865 was opened to track re-evaluation of the licensees ISFSI pad and subsequent resolution of the issue. This URI is further discussed in Section 4OA7, Licensee Identified Violations of this report.

The licensee revised calculations 2.4.3 BYR-08-027 and performed 2.4.3 BYR-10-095, a new calculation, as a result of inspector questioning associated with URI 07200068/2009001-01 and URI 07200068/2009001-02. Region III staff requested assistance through a technical assistance request of the Division of Spent Fuel Storage and Transportation, to review the two revised analyses to determine if the licensees evaluations met regulatory requirements.

b. Findings

Introduction The inspectors identified a Severity Level IV NCV of 10 CFR 72.212 (b)(2)(i)(B),

Conditions of a General License Issued under 72.210. Specifically, the inspectors identified four examples where the licensee failed to design the ISFSI pad to adequately support the static and dynamic loads of the stored casks, considering potential amplification of earthquakes through soil-structure interaction (SSI).

Description:

The inspectors identified four examples where the licensee failed to meet the requirements of 10 CFR 72.212 (b)(2)(i)(B).

1. The ISFSI must be designed to adequately support the static and dynamic loads considering potential amplification of earthquakes through SSI as required by 10 CFR 72.212. The dynamic analysis presented did not capture 3-D effects, such as torsion, due to a partially loaded pad. An asymmetrically loaded pad will have a torsional dynamic response, and it is anticipated that acceleration in the short direction will be lower for a fully loaded symmetric structure than for the partially loaded non-symmetric structure. The licensee failed to analyze the pad for the worst case cask configuration on the ISFSI pad and thus failed to adequately address increased torsional dynamic responses on the ISFSI pad. This example was previously documented by URI 07200068/2009001-01 Pad Structural Evaluation Revisions to Address Partial Sequential Loading."

2. The ISFSI must be designed to adequately support the static and dynamic loads

considering potential amplification of earthquakes through SSI as required by 10 CFR 72.212. In the design basis dynamic analysis of the Byron ISFSI pad the methodology used to develop the SSI model and ensuing SSI analyses used best estimate soil properties.

American Society of Civil Engineers (ASCE) Standard 4-98 (Ref. 6) Section 3.3.1.7 states the following: The uncertainties in the SSI analysis shall be considered. In lieu of a probabilistic evaluation of uncertainties, an acceptable method to account for uncertainties in SSI analysis is to vary the low strain soil shear modulus. Low strain soil shear modulus shall be varied between the best estimate value times (1+Cv) and the best estimate value divided by (1+Cv), where Cv is a factor that accounts for uncertainty in the SSI analysis and soil properties. If sufficient, adequate soil investigation data are available, the mean and standard deviation of the low strain shear modulus shall be established for every soil layer. The Cv shall be established so that it will cover the mean plus or minus one standard deviation for every layer. The minimum value of Cv shall be 0.5. When insufficient data are available to address uncertainties in soil properties, Cv shall be taken as no less than 1.0.

The licensee used, in part, ASCE 4-98 as industry guidance for completion of the SSI. However, the licensee failed to address uncertainties in the soil in accordance with this standard. Discussions with DSFST staff have determined that this omission is non-conservative. The omission will reduce the licensees calculated safety factor and should have been included in the licensees analysis.

3. The ISFSI must be designed to adequately support the static and dynamic loads

considering potential amplification of earthquakes through SSI as required by 10 CFR 72.212.

In the licensees SSI model the bedrock outcrop (which is also the base of the SSI model) was modeled as a fixed mass, and therefore was unable to move and transmit seismic waves. The earthquake control motions were therefore applied as an inertia force time history to each mass: cask center of gravity, pad center of gravity, and soil mass center of gravity. This methodology is non-physical. The inspectors recognized that this non-physical methodology may be theoretically correct for a linear analysis; however, the inspectors had no evidence that this methodology was applicable to a non-linear problem wherein a cask is allowed to slide, tip or lose complete contact with the pad. The inspectors noted that in every known SSI methodology that had been reviewed and approved by the NRC, the control motion was applied at a bedrock outcrop or comparable soil layer. This was physically how the earthquake ground motion arrived at the site. The seismic waves arrive at the bedrock outcrop, were filtered and amplified by the soil layers between the rock outcrop and the ground surface and generated motion to the ISFSI pad.

The licensee did not provide adequate justification and documentation for use of a new SSI analysis methodology.

4. The ISFSI must be designed to adequately support the static and dynamic loads

considering potential amplification of earthquakes through SSI as required by 10 CFR 72.212. In the licensees analysis, a single set of three dimensional (two horizontal and one vertical) acceleration time-histories were developed to envelop the 5% damped Regulatory Guide 1.60 response spectra to perform the non-linear SSI analysis. The use of a single set of 3-D time-histories was not standard practice for performing a non-linear SSI analysis. ASCE 4-98 Section 3.2.2.3(d) "Nonlinear Analysis" states the following: "In general, more than one set of acceleration time-histories, meeting the requirements of Section 2.3, should be used, and the results of the analyses shall be averaged. NUREG/CR-6865 also discusses this same issue and states the following in Section 4.1: ...the seismic response of a dry cask using one time-history might not always lead to a predictable response. It is increasingly obvious that a suite of earthquake inputs should be examined in order to obtain statistically stable mean and standard variation in the response to form the basis for design decision. This would require multiple runs using several earthquake records.

The NUREG further provided evidence that the difference in maximum response among 5 sets of time histories varied by as much as a factor of 6 for the same spectral shape. This showed that the effect of the differences in frequency content and phasing within the 5 sets of time-histories had a significant influence on response. Due to the potentially large differences in response that could result from using different earthquake time-histories as input to a non-linear SSI analysis, the inspectors determined that the licensees use of only a single set of acceleration time-histories to perform a non-linear SSI analysis may have significantly underestimated the predicted seismic response and thus did not conservatively meet the requirements of 10 CFR 72.212.

As an immediate corrective action and given the need for the licensee to load ISFSI casks and move them onto the pad, the licensee restricted the total load applied to the ISFSI pad by allowing a maximum of ten casks and then only at every other cask location in each direction on the pad, so that for any cask on the pad an open (unused)location would be adjacent to it in both the length and width directions of the pad.

Because this restriction on the number of casks and loading pattern significantly reduces the total load distribution on the pad, the licensee concluded that for this reduced loading the concrete pad could adequately support the static and dynamic loads.

Analysis The inspectors determined that the previously discussed examples were a violation of more than minor significance. The violation was determined to be more than minor because if left uncorrected, a failure of the ISFSI pad could lead to a more significant safety concern. Consistent with the guidance in Section 2.2 of the NRC Enforcement Manual, ISFSIs are not subject to the SDP and, thus, traditional enforcement will be used for these facilities. Consistent with the guidance in Section 2.6.D of the NRC Enforcement Manual, if a violation does not fit an example in the Enforcement Policy Violation Examples, it should be assigned a severity level:

(1) Commensurate with its safety significance; and

(2) informed by similar violations addressed in the Violation Examples. The inspectors determined that the violation could be screened using a similar violation example discussed in Section 6.5.d.1 of the NRC enforcement policy.

Based on this the violation was determined to be of very low safety significance (Severity Level IV).

Enforcement Title 10 CFR 72.212 (b)(2)(i)(B) requires, in part, that the licensee perform written evaluations prior to use, that establish that the cask storage pads and areas have been designed to adequately support the static and dynamic loads of the stored casks, considering potential amplification of earthquakes.

Contrary to the above, the licensees completed evaluation did not adequately evaluate the cask storage pad to support static and dynamics loads of the stored casks considering potential amplification of earthquakes. This is a violation of 10 CFR 72.212 (b)(2)(i)(B), Conditions of a General License Issued under 72.210.

Because this matter was of very low safety-significance (Severity Level IV), and has been entered into the licensees corrective action program (IR 957945 and IR 1102835),this violation is being treated as a NCV consistent with the NRC Enforcement Policy.

(NCV 07200068/2010001-01). This closes Unresolved Item (URIs 07200068/2009001-01;05000454/2009009-01; 05000455/2009009-01).

4OA6 Management Meetings

Interim Exit Meetings Interim exits were conducted for:

The results of the Emergency Preparedness program inspection with Mr. B. Adams on July 16, 2010.

Radiation monitoring instrumentation protective equipment and verification inventories of materials tracked in the National Source Tracking System with Mr. B. Adams on July 16, 2010.

The results of the ISFSI dry run readiness inspections with Mr. B. Adams on August 23, 2010.

The inspectors confirmed that none of the potential report input discussed was considered proprietary.

4OA7 Licensee Identified Violations

The following violations of very low safety significance (Green) were identified by the licensee and were violations of the NRC requirements which meets the criteria of Section 2.3.2 of the NRC Enforcement Policy, NUREG 1600, for being dispositioned as NCVs.

1) Title 10 CFR 50, Appendix B, Criterion III, Design Control states in part that the design control measures shall provide for verifying and checking adequacy of the design, such as by design review. The FHB crane was designed to Seismic Category I requirements. The licensee used compliance with ASME NOG-1-2004 as the basis for its upgrade to single failure proof. Section 4153.6 of ASME NOG-1-2004 describes the required boundary conditions to be used in the crane analysis.

Contrary to above, on July 7, 2009, in Calculation 36272-17, Revision 11, the licensees design control measures failed to verify that the boundary conditions as required by ASME NOG-1-2004 were applied in the evaluation. Specifically, determination of seismic loads based on coefficient of friction was not consistent with the required crane boundary conditions per Section 4153.6 of ASME NOG-1-2004.

This violation was of very low safety significance because the crane was not yet in operation as a single failure proof crane. The inspectors reviewed the revised calculation. The licensee entered this issue into its corrective action program (IR 966184). This licensee-identified violation closes URIs 005000454/2009004-04;05000455/2009004-04.

2) Title 10 CFR 72.212 (b)(2)(i)(B) requires, in part, that the licensee perform written evaluations prior to use, that establish that the cask storage pads and areas have been designed to adequately support the static and dynamic loads of the stored casks, considering potential amplification of earthquakes. The licensee generated Calculation 2.4.3-BYR08-30 to satisfy the requirements of 10 CFR 72.212 (b)(2)(i)(B).

Contrary to the above, on September 18, 2009, in Calculation 2.4.3-BYR08-30, in lieu of performing a detailed dynamic analysis to determine seismic response of the cask, the licensee used the methodology described in the NUREG/CR6865, Parametric Evaluation of Seismic Behavior of Free Standing Spent Fuel Dry Cask Storage System. The licensee determined that the calculation contained a number of assumptions and did not demonstrate the Byron ISFSI pad was bounded by the analyzed pad in NUREG/CR-6865. The licensee revised their calculation and performed a soil structure interaction analysis to address the concern and satisfy the requirements of 10 CFR 72.212 (b)(2)(i)(B). The inspectors reviewed the revised calculation. The licensee entered this issue into its corrective action program (IR 966751). This licensee-identified violation closes URIs 07200068/2009001-02;05000454/2009009-02; and 05000455/2009009/02.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Adams, Plant Manager

K. Anderson, Radiation Protection

A. Daniels, Nuclear Oversight

D. Drawbaugh, Emergency Preparedness Manager

C. Gayheart, Operations Director

B. Kartheiser, Emergency Preparedness Coordinator

J. Langan, Licensing Engineer

S. Merrill, Corporate Emergency Preparedness

J. Reed, Radiation Protection

D. Spitizer, Regulatory Assurance

D. Enright, Site Vice President

B. Askren, Security Director

D. Gudger, Regulatory Assurance Manager

T. Hulbert, Regulatory assurance NRC Coordinator

P. Johnson, Lead NOS Assessor

S. Kerr, Chemistry Manager

R. Lloyd, Performance Improvement Manager

T. Leaf, Operations SOS

J. Feimster, Maintenance and Technical Training Manager

B. Spahr, Maintenance Director

E. Hernandez, Engineering Director

B. Barton, Radiation Protection Manager

Nuclear Regulatory Commission

E. Duncan, Chief, Reactor Projects Branch 3

R. Skokowski, Chief, Plant Support Branch

J. Robbins, Resident Inspector

Attachment

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened

05000454/2010004-01 NCV Failure to Perform Adequate Evaluation for Crane Upgrade

05000455/2010004-01 (Section 40A5)

05000454/2010004-02 URI Relying on the Use of Friction as a Restraining Force when

05000455/2010004-02 the HI-TRAC is Placed on top of the HI-STORM (Section 40A5)

200068/2010001-01 NCV Failure to Design the ISFSI Pad to Adequately Support the Static and Dynamic Loads of Stored Casks (Section 40A5)

Closed

05000454/2010004-01 NCV Failure to Perform Adequate Evaluation for Crane Upgrade

05000455/2010004-01 (Section 4OA5)

200068/2010001-01 NCV Failure to Design the ISFSI Pad to Adequately Support the Static and Dynamic Loads of Stored Casks (Section 4OA5)

05000454/2009004-04 URI Use of Friction in Design of FHB Crane to Single Failure

05000455/2009004-04 Proof (Section 4OA7)

05000454/2009004-05 URI Unresolved Technical Concerns on Design of Seismic

05000455/2009004-05 Restraint on FHB Crane Trolley (Section 4OA5)

05000454/2009009-01 URI Pad Structural Evaluation Revisions to Address Partial

05000455/2009009-01 Sequential Loading (Section 4OA5)

200068/2009001-01

05000454/2009009-02 URI Use of NUREG/CR6865 Results in Lieu of Dynamic

05000455/2009009-02 Analysis (Section 4OA7)

200068/2009001-02

05000455/2009-001-01 LER Late Entry Into TS Condition Associated with Reactor Coolant System Leakage Characterization Resulting in a condition Prohibited by TSS (Section 4OA3)

Discussed

None Attachment

LIST OF DOCUMENTS REVIEWED