IR 05000277/2008004

IR 05000277-08-004, 05000278-08-004; on 07/01/2008 - 09/30/2008; Peach Bottom Atomic Power Station (Pbaps), Units 2 and 3; Operability Evaluations; Event Follow-up; Public Radiation Safety

ML083190078
Person / Time
Site:	Peach Bottom
Issue date:	11/13/2008
From:	Paul Krohn Reactor Projects Region 1 Branch 4
To:	Pardee C AmerGen Energy Co, Exelon Nuclear
KROHN, PG
References
IR-08-004
Download: ML083190078 (42)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ber 13, 2008

SUBJECT:

PEACH BOTTOM ATOMIC POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000277/2008004 and 05000278/2008004

Dear Mr. Pardee:

On September 30, 2008, the United States Nuclear Regulatory Commission (NRC) completed an inspection at your Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The enclosed integrated inspection report documents the inspection results, which were discussed on October 17, 2008, with Mr. William Maguire 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.

The report documents one self-revealing finding, one inspector-identified finding, and one Severity Level IV violation. Two of these findings are of very low safety significance (Green)

and involve violations of NRC requirements. However, because of the very low safety significance and because these three findings have been entered into your corrective action program (CAP), the NRC is treating these findings as a non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCVs in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the PBAPS.

In accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390 of the NRC's

"Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the 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/

Paul G. Krohn, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket Nos.: 50-277, 50-278 License Nos.: DPR-44, DPR-56

Enclosures:

Inspection Report 05000277/2008004 and 0500278/2008004 w/Attachment: Supplemental Information

REGION I==

Docket Nos.: 50-277, 50-278 License Nos.: DPR-44, DPR-56 Report No.: 05000277/2008004 and 05000278/2008004 Licensee: Exelon Generation Company, LLC Facility: Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Location: Delta, Pennsylvania Dates: July 1, 2008 through September 30, 2008 Inspectors: F. Bower, Senior Resident Inspector M. Brown, Resident Inspector J. DAntonio, Senior Operations Engineer D. Kern, Senior Resident Inspector, Three Mile Island (TMI)

R. Nimitz, Senior Health Physicist T. OHara, Reactor Inspector J. Tifft, Reactor Inspector Approved by: Paul G. Krohn, Chief Reactor Projects Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000277/2008004, 05000278/2008004; 07/01/2008 - 09/30/2008; Peach Bottom Atomic

Power Station (PBAPS), Units 2 and 3; Operability Evaluations; Event Follow-up; Public Radiation Safety.

The report covered a three-month period of inspection by resident inspectors and announced inspections by a regional senior health physicist, two regional reactor inspectors, a senior operations engineer, and a senior resident inspector (SRI) from Three Mile Island. One Green self-revealing NCV, one Green inspector-identified NCV, and one Severity Level IV NCV were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or 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.

Cornerstone: Mitigating Systems

Green.

A self-revealing NCV of 10 CFR Part 50, Appendix B, Criteria V, Instructions,

Procedures and Drawings, was identified for a failure to follow procedure, WC-PB-2000,

Outage Control Center (OCC) Emergent Issue Response, that resulted in an inadequate extend of condition (EOC) evaluation being performed for an emergency service water (ESW) leak that was discovered on the E-1 emergency diesel generator (EDG). Specifically, Operations personnel failed to look at similar ESW locations on the E-2, E-3, and E-4 EDGs. This resulted in a nine-day delay in discovering a similar leak on the E-4 EDG.

This finding is greater than minor because it is similar to the example 4a., Insignificant Procedural Errors, in Manual Chapter 0612, Appendix E, in that, the later evaluation of the ESW leak discovered on the E-4 EDG resulted in safety-related equipment being adversely affected. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the finding was of very low safety significance (Green) since it did not represent an actual loss of system safety function for the ESW system. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution (PI&R), Corrective Action Program (CAP) because the licensee failed to thoroughly evaluate the EOC of the leak on the E-1 EDG and it resulted in a nine-day delay in discovering additional leaks associated with the E-3 and E-4 EDGs. [IMC 0305, aspect P.1(c). (Section 1R15)

Severity Level IV. The inspectors identified a NCV of 10 CFR 50.72(b)(2)(xi) because the NRC Operations Center was not notified of a reportable event. Specifically, PBAPS did not formally report, to the NRC Operations Center, a planned press release and the notification of other government agencies regarding a transformer fire and petroleum product spill event that occurred on July 23 and 24, 2008.

This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. This event was related to pubic health and safety because it involved a fire emergency that contributed to the loss of two of the plants three offsite power sources. This event was related to protection of the environment because it involved the spill of a more than minor quantity of oil that required reporting to the State of Pennsylvania. While reviewing this finding, the inspectors considered the fact that the NRC was informally notified. The inspectors considered the above and evaluated the severity of this violation using the criteria contained in Supplement I - Reactor Operations and Section VI.A.1 of the NRCs Enforcement Policy and determined that this finding met the criteria for disposition as a NCV. (Section 4OA3.1)

Cornerstone: Public Radiation Safety

Green.

The inspectors identified a NCV of 10 CFR 20, Appendix G, Section III.C. 5.

Specifically, the licensee did not conduct a Quality Assurance Program sufficient to assure conformance with 10 CFR 61.55, in that, the program was not adequate to identify incorrect gamma spectroscopy analyses of a principal gamma emitting radionuclide used to scale hard-to detect radionuclides for purposes of waste classification in accordance with 10 CFR 61.55. The licensee entered the deficiency into its CAP (IR799894).

The failure to conduct a sufficiently robust 10 CFR 61 Quality Assurance Program, to assure conformance with 10 CFR 61.55, is a performance deficiency that was reasonably within the licensees ability to foresee and correct, and which should have been prevented. The finding is more than minor because it affected the associated cornerstone objective in that the licensees 10 CFR 61 Quality Assurance Program did not identify incorrectly analyzed waste samples used to classify radioactive waste for land disposal. This finding was determined to be of very low safety significance because no radiation limits were exceeded, there was no breach of packaging, there was no certificate of compliance finding, there was no low level burial ground non-conformance, and lastly, there was no failure to make notifications or provide emergency notification information. The cause of this finding was related to the cross-cutting area of PI&R, self and independent assessments component, in that, although actions were taken to coordinate and communicate results from assessments to affected personnel, corrective actions were not sufficiently comprehensive to identify incorrect vendor analyses [IMC 0305, aspect P.3(c). (Section 2PS2)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Unit 2 began the inspection period at 100 percent rated thermal power (RTP). On July 8, 2008, the unit began the end-of-cycle coastdown. On September 14, 2008, the unit was shutdown to start its 17th refueling outage (RFO) (2PR17) and remained in the outage through the end of the inspection period.

Unit 3 began the inspection period at 100 percent RTP where it remained until the end of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - 1 System Sample)

.1 Impending Adverse Weather Conditions

a. Inspection Scope

Since Tropical Storm Hanna was projected to pass over PBAPS on September 6, 2008, the inspectors reviewed PBAPSs overall preparations for the expected high winds and heavy rain. On September 5, 2008, the inspectors walked down portions of the ESW system, the EDGs, and the North Substation. These systems were selected because their safety-related functions could be affected by adverse weather. The inspectors observed plant conditions, evaluating those conditions using criteria documented in OP-PB-108-111-1001, Preparation for Severe Weather, Revision 3. The inspectors also toured the plant grounds for loose debris which could become missiles during the expected high winds, and verified operators could access controls and indications for those systems required for safe control of the plant.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (7111104Q - 3 Partial Samples)

.1 Partial Walkdown

a. Inspection Scope

The inspectors performed a partial walkdown of three systems to verify the operability of redundant or diverse trains and components when safety-related equipment was inoperable. The inspectors performed walkdowns to identify any discrepancies that could impact the function of the system and potentially increase risk. The inspectors reviewed applicable operating procedures, walked down system components, and verified that selected breakers, valves, and support equipment were in the correct position to support system operation. The three systems reviewed were:

• A ESW System with B ESW Out-of-Service (OOS);

• 2 A Core Spray (CS) System with 2 B CS System OOS; and

• Unit 3 Reactor Core Isolation Cooling (RCIC) with Unit 3 High Pressure Coolant Injection (HPCI) OOS.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05 - 5 Samples)

.1 Fire Protection - Tours

a. Inspection Scope

The inspectors reviewed PBAPSs Fire Protection Plan, Technical Requirements Manual (TRM), and the respective pre-fire action plan procedures to determine the required fire protection design features, fire area boundaries, and combustible loading requirements for the areas examined during this inspection. The fire risk analysis was reviewed to gain risk insights regarding the areas selected for inspection. The inspectors performed walkdowns of five areas to assess the material condition of active and passive fire protection systems and features. The inspection was also performed to verify the adequacy of the control of transient combustible material and ignition sources, the condition of manual firefighting equipment, fire barriers, and the status of any related compensatory measures. The following five fire areas were reviewed for impaired fire protection features:

• Unit 3 - 3 A and 3 C Residual Heat Removal (RHR) Pump Room and Heat Exchanger Room, 916 and 116 Elevation (Fire Zone 11 and 12A);

• Unit 2 Refuel Floor, 234 Elevation (Fire Zone 57);

• Unit 2 Reactor Building, 135 Elevation (Fire Zone 5H and 5P);

• Unit 3 Reactor Building Closed-Cooling Water Room, 116 Elevation (Fire Zone 12B); and

• Unit 3 Main Transformer Yard (Fire Zone 152).

b. Findings

No findings of significance were identified.

1R06 Flood Protection (71111.06 - 1 Sample)

a. Inspection Scope

The inspectors reviewed selected risk-important plant design features intended to protect the plant and its safety-related equipment from internal flooding events. The inspectors reviewed the flood analysis and Updated Final Safety Analysis Report (UFSAR). The inspectors walked down the Unit 2 RCIC pump room to evaluate the condition of penetration seals, watertight doors, and other internal design features to verify that they were as described in the Individual Plant Examination (IPE).

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection (ISI) (71111.08 - 1 Sample)

a. Inspection Scope

The purpose of this inspection was to assess the effectiveness of the licensee=s ISI program for monitoring degradation of the reactor coolant system boundary, risk significant piping system boundaries, and the containment boundary. The inspectors assessed the ISI activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI and applicable NRC regulatory requirements.

The inspectors selected a sample of nondestructive examination (NDE) activities for observation or review, and evaluation for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of activities associated with the repair/replacement of safety-related pressure boundary components. The sample selection was based on the inspection procedure (IP) objectives, risk significance, and availability. Specifically, the inspectors focused on components and systems where degradation would result in a significant challenge to the integrity of pressure boundary components.

The inspectors performed an observation of one volumetric examination (ultrasonic) and portions of a surface examination (liquid penetrant). In addition, the inspectors performed a documentation review of a magnetic particle surface examination. The sample selection included the following:

• Ultrasonic Test (UT), volumetric examination, of weld # CH-MB reactor head meridonal weld. The examination report for this component weld contained several indications which were all accepted as being less than the error band and were attributed to normal ultrasonic variation. A magnetic particle examination was performed on the CH-MB inside surface. No recordable magnetic particle testing indications were found.

• UT, volumetric examination, of the nozzle-to-safe-end weld of component 2-AS-1 nozzle.

• UT, volumetric examination, of the nozzle-to-safe-end weld of component nozzle N5A.

The inspectors performed an evaluation of work activities during a drywell entry this inspection period and, observed the examinations performed for the first and third samples above. The inspectors reviewed documentation from the current outage inspection of the containment conducted in accordance with MA-PB-793-001, Revision 1, Visual Examination of Containment Vessels and Internals.

The inspectors reviewed inspection reports for the remote visual examination of the Steam Separator Lower Bolt Ring Gussets. The inspectors reviewed vendor report 0000-0091-3414-R1 which noted the indications reported in 2008 and compared the present condition with the previous inspection results from 2006. Condition Reports (CRs) were placed in their CAP for engineering evaluation and disposition.

The inspectors selected a sample of repair/rework activities for review which required the development and implementation of an ASME Section XI repair plan. The inspectors reviewed documentation for the weld repair of two pressure boundary valves in ASME Class 3 systems. The inspectors reviewed the ASME Section XI plans, work scope, activity sequence, weld filler metal selection, weld procedure specifications and procedure qualification records, welder qualifications, specified non-destructive tests, acceptance criteria, and post-work testing. The following samples were inspected:

• Action Request (AR) A1587644 (MO-2-23-031) dealt with the repair/replacement of the HPCI flush piping to the torus. The repair consisted of the mechanical removal of eroded piping, fabrication of replacement piping and installation of new piping sections. Additionally, the work order (WO) contained the results of the post-installation testing conducted on the replaced piping.

• (AR A1602845 (HV-0-33-510) was initiated to facilitate the removal, testing, rebuilding, inspection and welding of leaking ESW piping to the Unit 2 EDG coolers. Acceptance testing of the completed repair and welding was specified in the repair plan. A visual examination was specified for the installation welds and a system pressure test was specified to verify system integrity.

No sample of a previously identified recordable indication, accepted as-is (for continued service from the previous and the current outage), was available for review during the inspection.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11Q - 1 Sample)

.1 Resident Inspector Quarterly Review

a. Inspection Scope

On August 26, 2008, the inspectors observed operators in PBAPSs simulator during licensed operator requalification training to verify that operator performance was adequate and that evaluators were identifying and documenting crew performance issues. The inspectors verified that performance issues were discussed in the crews post-scenario critiques. The inspectors discussed the training, simulator scenarios, and critiques with the operators, shift supervision, and the training instructors. The evaluated scenario observed for this one sample involved the exercising of the below listed procedures:

• ON-118, Loss of Turbine Building Closed Cooling Water System, Revision 5;

• ON-119, Loss of Instrument Air, Revision 15;

• SE-11, Loss of Off-site Power; and

• T-101, Reactor Pressure Vessel Control, Revision 19.

b. Findings

No findings of significance were identified.

.2 Senior Reactor Operator Limited to Fuel Handling Requalification Program (71111.11B -

1 Sample)

a. Inspection Scope

The requalification program for Limited Senior Reactor Operator (LSRO) Limited to Fuel Handling was evaluated using NUREG 1021, Revision 9, Operator Licensing Examination Standards for Power Reactors, and IP Attachment 7111111, Licensed Operator Requalification Program.

A review was conducted of recent operating history documentation regarding fuel handling found in inspection reports, licensee event reports, the licensee=s CAP, and the most recent NRC plant issues matrix. The inspectors also reviewed specific events from the licensee=s CAP to determine if possible training deficiencies existed.

The inspectors evaluated the Peach Bottom 2007 operating tests and the Peach Bottom 2006 written examinations for quality and compliance with the Examiner=s Standards.

On August 12, 2008, the results of the biennial written examination and annual operating tests for 2008 were reviewed to determine whether pass/fail rates were consistent with the guidance of NUREG-1021, Revision 9, Operator Licensing Examination Standards for Power Reactors. Performance of all individuals over two years was reviewed to check for adverse trends.

Two years of records for requalification training attendance and license reactivation for all four LSROs were reviewed for compliance with license conditions and NRC regulations. Medical records for three individuals were reviewed.

A sampling of feedback was reviewed and training materials were evaluated for response to this feedback. These materials were also reviewed for incorporation of plant modifications and industry events.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12 - 2 Samples)

a. Inspection Scope

The inspectors evaluated PBAPSs work practices and follow-up corrective actions for structures, systems, and components (SSCs) and identified issues to assess the effectiveness of PBAPSs maintenance activities. The inspectors reviewed the performance history of SSCs and assessed Exelons EOC determinations for those issues with potential common cause or generic implications to evaluate the adequacy of the PBAPSs corrective actions. The inspectors assessed PBAPSs PI&R actions for these issues to evaluate whether PBAPS had appropriately monitored, evaluated, and dispositioned the issues in accordance with Exelon procedures, including ER-AA-310, Implementation of the Maintenance Rule, and the requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance. In addition, the inspectors reviewed selected SSC classifications, performance criteria and goals, and Exelons corrective actions that were taken or planned, to evaluate whether the actions were reasonable and appropriate. The inspectors performed the following two samples:

• Cause for Circuit Breaker 3435 Trip During No. 1 Transformer Failure Unknown (IR 811332); and

• Perform a Root Cause Analysis for the ESW Piping Leakage Issues (IR 798807).

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 4 Samples)

a. Inspection Scope

The inspectors evaluated PBAPS=s implementation of their Maintenance Risk Program with respect to the effectiveness of risk assessments performed for maintenance activities that were conducted on SSCs. The inspectors also verified that the licensee managed the risk in accordance with 10 CFR Part 50.65(a)(4) and procedure WC-AA-101, On-line Work Control Process. The inspectors evaluated whether PBAPS had taken the necessary steps to plan and control emergent work activities and to manage overall plant risk. The inspectors selectively reviewed PBAPSs use of the Paragon online risk monitoring software, and daily work schedules. The activities selected were based on plant maintenance schedules and systems that contributed to risk. The inspectors completed four evaluations of maintenance activities on the following:

• HV-0-33-504D, Replace Valve and Elbows (Work Order (WO) C0225717);

• CHK-2-32-502C, Has a Through Wall Leak (WO C0225614);

• Open Unit 3 Output Breaker (CB-65) to Support Limerick Experiencing Hot Spot on 5010 Line Disconnect (AR A1675657); and

• Repair\Replace Degraded ESW Piping and Valve (HV-0-33-504C) to E-3 EDG (WO C0225784).

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - 5 Samples)

a. Inspection Scope

The inspectors reviewed five issues to assess the technical adequacy of the operability evaluations, the use and control of compensatory measures, and compliance with the licensing and design bases. Associated adverse condition monitoring plans, engineering technical evaluations, and operational and technical decision making documents were also reviewed. The inspectors used Technical Specification (TS), TRM, UFSAR, and associated Design Basis Documents as references during these reviews. The issues reviewed included:

• Failure to Meet Initial System Parameters During Unit 2 HPCI Surveillance Testing (IR 785568);

• Suspected ESW Leak/Pipe Degradation on E-1 EDG (IR 793791);

• 2 C Traversing Incore Probe Detector Cannot be Withdrawn (IR 813156);

• OP-Evaluation 08-005, Thru Wall Leakage was Identified in the B ESW Supply Line to the E-3 and E-4 EDGs (IR 796776); and

• HV-0-33-504D, Valve Body Leak - Generic Letter (GL) 90-05 EOC Requirements.

b. Findings

Introduction:

A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures and Drawings, was identified for a failure to follow procedure WC-PB-2000, OCC [Outage Control Center] Emergent Issue Response, that resulted in an inadequate EOC evaluation being performed for an ESW leak that was discovered on the E-1 EDG.

Description:

On July 6, 2008, while performing a hydrostatic test on the high pressure service water system with the ESW system in standby mode under system pressure, a leak was discovered on an elbow on the upstream side of the B supply header to the E-1 EDG. Per TRM 3.10, Condition B, any ASME Class 2 or 3 moderate energy piping pressure boundary leakage must have an initial operability determination (OD)completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In addition, if the initial OD supports operability, then engineering must evaluate and characterize the flaw in an engineering analysis performed, as soon as possible, but not to exceed a maximum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of discovery.

In response to this event, on July 6, 2008, the OCC was activated, staffed, and procedure WC-PB-2000, OCC Emergent Issue Response, Revision 2 was entered.

Step 4.1.3.6 of this procedure requires an EOC review be performed on redundant equipment, the opposite unit, etc. The inspectors concluded that the redundant equipment would have been the elbows on the A and B ESW supply headers to the E-2, E-3, and E-4 EDGs. In order to look at these locations, the diamond plate flooring needed to be removed. This was not immediately done on July 6. The EOC review consisted of entering the respective EDG rooms and looking at the accessible ESW piping. No additional leaks or degraded areas were identified.

On July 7th, Ultrasonic Testing (UT) was conducted on ESW piping in the E-1 EDG room and determined that pipe wall thickness around the leak location was less than Code requirements. On July 8th, Operability Evaluation 08-004 concluded that the system was degraded but operable and directed that repairs be completed no later than July 31, 2008. The use of guidance in GL 90-05 required that follow-up augmented inspections of five other ESW pipe locations be completed by July 21, 2008.

On July 15th while preparing the ESW piping to conduct the augmented UT inspections, several degraded pipe areas and one degraded valve were identified under the diamond plate flooring in the E-3 and E-4 EDG rooms. The valve leak initially resulted in the both trains of ESW being declared inoperable and the plant entering a 12-hour shutdown limiting condition for operation (LCO), once the leak was isolated the E-4 EDG became inoperable and the plant entered a 14 day shutdown LCO.

The inspectors concluded that the performance deficiency was an inadequate immediate EOC evaluation required by procedure WC-PB-2000, OCC Emergent Issue Response, on July 6, 2008, which led to a failure to detect additional degraded ESW piping.

Analysis:

The finding was more than minor because it is similar to the example 4a.,

Insignificant Procedural Errors, in Manual Chapter 0612, Appendix E, because the augmented inspections performed resulted in safety equipment being declared inoperable.

Using the Phase 1 worksheet in Manual Chapter 0609, SDP, the finding was of very low safety significance (Green) since it did not represent an actual loss of system safety function for the ESW system. This finding has a cross-cutting aspect in the area of PI&R, CAP, because the licensee failed to thoroughly evaluate the EOC of the leak on the E-1 EDG and it resulted in a nine-day delay in discovering additional leaks associated with the E-3 and E-4 EDGs. (Aspect P.1(c)).

Enforcement:

10 CFR 50 Appendix B, Criteria V, Instructions, Procedures and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with those procedures. Procedure WC-PB-2000, OCC Emergent Issue Response, Revision 2, Step 4.1.3.6 requires that during event investigation an EOC review be conducted for redundant equipment. Contrary to this, between July 6, 2008 and July 15, 2008, the licensee failed to look at redundant equipment when a leak was discovered on an elbow on the upstream side of the B supply header to the E-1 EDG. As a result, identification of additional leaks associated with the E-3 and E-4 EDGs was delayed. Because the finding is of very low safety significance and has been entered into PBAPSs CAP (IR 798033), this violation is being treated as a Green NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000277; 278/2008004-01, Inadequate EOC Review Results in Delay in Discovery of ESW Leaks.

1R18 Plant Modifications (71111.18 - 3 Samples)

.1 Permanent Plant Modifications (1 Sample)

a. Inspection Scope

The inspectors observed selected ongoing and completed work activities to implement a design change that affected both units, while PBAPS Units 2 and 3 were online. The review was conducted to verify that the installation was consistent with the design control documents. The inspectors also reviewed procedures CC-AA-102, "Design Input and Configuration Change Impact," and CC-AA-103, "Configuration Change Control,"

and reviewed PBAPSs engineering change request (ECR) package that approved a design change to use an alternative valve to replace hand valve HV-0-33-504D, the B train ESW inlet isolation valve to the E-4 EDG coolers. ECR PB 08-00287, HV-0-33-504D Needs Replacement Due to Leakage, was developed to compare the critical characteristics of the original valve and the alternative valve model from a different manufacturer. The review was conducted to verify that the design bases, licensing bases, and performance capability of safety-related systems and components were not degraded through this modification.

b. Findings

No findings of significance were identified.

.2 Temporary Plant Modifications (2 Samples)

a. Inspection Scope

The inspectors reviewed two temporary modifications to verify that implementation of the modifications did not place the plant in an unsafe condition. The review was also conducted to verify that the design bases, licensing bases, and performance capability of risk significant SSCs had not been degraded as a result of these modifications. The inspectors verified the modifications were developed, planned, and implemented in accordance with the requirements of CC-AA-112, Temporary Configuration Changes, Revision 12 (IR 822365). The inspectors also verified the modified equipment alignment through control room instrumentation observations; UFSAR, drawings, procedures, and WO reviews; and plant walkdowns of accessible equipment deficiencies. The following temporary modifications were reviewed:

• ECR 08-00229, Install Water Mister on Main Transformer; and

• WO C0225542, Install Thermocouple Upstream of Main Steam Line A.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - 5 Samples)

a. Inspection Scope

The inspectors observed selected portions of post-maintenance testing (PMT) activities and reviewed completed test records. The inspectors observed whether the tests were performed in accordance with the approved procedures and assessed the adequacy of the test methodology based on the scope of maintenance work performed. In addition, the inspectors assessed the test acceptance criteria to evaluate whether the test demonstrated that the tested components satisfied the applicable design and licensing bases and the TS requirements. The inspectors reviewed the recorded test data to verify that the acceptance criteria were satisfied. The inspectors reviewed five PMTs performed in conjunction with the following maintenance activities:

• Unit 3 HPCI - Auxiliary Oil Pump Breaker Maintenance (R0863509);

• Replace Diesel-Driven Fire Pump (C0224289);

• Unit 2 A Drywell Floor Drain Sump Pump Coupling Inspection (R0977757);

• Unit 2 E-12 Bus Transfer Relay (2-54-102-1501(8)) and Diesel Generator (DG) Start Relay (2-54-183-1501(8) Calibrations (R0975818); and

• Unit 2 - CS Inboard Discharge Valve Maintenance (R1047430).

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

.1 Peach Bottom Unit 2 RFO 17 (P2R17)

a. Inspection Scope

The Unit 2 RFO (P2R17) was conducted from September 14, 2008, through the end of the inspection period. Prior to the start of P2R17 on September 14, 2008, the inspectors reviewed the stations work schedule and the Outage Risk Assessment Management (ORAM) Plan against procedures OU-PB-104, "Shutdown Safety Management Program; OU-PB-104-1001, "Shutdown Risk Management for Outages; and OU-AA-103, "Shutdown Safety Management Program." The ORAM plan was reviewed to confirm that the PBAPS had appropriately considered risk, industry experience, and previous site specific problems in developing and implementing a plan that maintained shutdown safety defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored the activities listed below to verify PBAPS controls over the outage activities:

• Observed the control room operators removing the main generator from the grid, completing a soft shutdown of Unit 2, including stabilizing the plant in Mode 3;

• Observed selected plant cool down activities;

• Conducted an initial drywell walkdown to check for unidentified leakage or other discrepant conditions;

• Maintenance of secondary containment as required by TS;

• Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TS when taking equipment OOS;

• Monitoring of decay heat removal operations, including the spent fuel pool cooling system;

• Monitoring reactor water inventory controls including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss;

• Monitoring the status and configuration of electrical systems and switchyard activities to ensure that TS were met;

• Monitored activities that could affect reactivity;

• Monitored refueling activities, including fuel handling and fuel receipt inspections;

• Independently reviewed other core verification activities;

• Conducted a closeout walkdown and inspection of the torus to check for material deficiencies and to verify that all debris was removed; and

• Identification and resolution of problems related to RFO activities.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 5 Samples)

a. Inspection Scope

The inspectors reviewed and observed selected portions of selected surveillance tests (STs), and compared test data with established acceptance criteria to verify the systems demonstrated the capability of performing the intended safety functions. The inspectors also verified that the systems and components maintained operational readiness, met applicable TS requirements, and were capable of performing design basis functions.

The five STs reviewed and observed included:

• ST-C-095-884-2, Sampling Diesel Fuel Prior to Delivery to On-Site Storage Tanks;

• ST-O-052-313-2, E-3 DG Slow Start Full Load and IST [IST Sample];

• ST-O-033-400-2, ESW Valve CHK-2-33-514 IST [IST Sample];

• ST/LLRT 20.01A.02, Main Steam Isolation Valve Local Leak Rate Test

[Isolation Valve Sample]; and

• SI3L-2-72-A1FQ, Functional Test of Emergency Core Cooling System (ECCS)

A Compensated Trip System.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Controls (71121.01 - 18 Samples)

a. Inspection Scope

The inspectors reviewed selected activities and associated documentation in the below listed areas. The evaluation of Exelon=s performance in these areas was against criteria contained in 10 CFR 20, applicable TSs, and applicable Exelon procedures.

.1 Inspection Planning - Performance Indicators

The inspectors reviewed performance indicators (PIs) for the Occupational Exposure Cornerstone. The inspectors also discussed and reviewed current performance, relative to the indicators, with Exelon personnel.

.2 Plant Walkdowns, Radiation Work Permit (RWP) Reviews, and Jobs-in-Progress

Reviews The inspectors walked down selected radiological controlled areas and reviewed housekeeping, material conditions, posting, barricading, and access controls to radiological areas. The inspectors made selective independent ambient radiation level measurements to verify radiological conditions. The inspectors observed and selectively reviewed on-going outage work activities.

During the Unit 2 outage, the inspectors toured the drywell; entered the Unit 2 torus to observe diving activities; observed ongoing Unit 2 refueling and in-vessel work activities; observed main steam isolation valve work activities; observed in-service inspections; observed on-going moisture separator work; reviewed and directly observed control rod drive work activities; directly observed reactor in-core instrumentation removal activities; and reviewed spent fuel pool work, including in-vessel inspection activities. The inspectors also observed on-going turbine and condenser work. The inspectors reviewed radiation protection job coverage and radiation work permit implementation.

The inspectors verified adequacy of radiological controls including use of multiple dosimeters and re-positioning of dosimeters for work in radiation dose rate gradients.

The inspectors reviewed electronic dosimeter alarm set-points for adequacy and conformity with survey indications and plant policy. The inspectors reviewed use of electronic dosimeters for monitoring of workers in high radiation areas.

The inspectors reviewed and discussed internal dose assessments, since the previous inspection, to identify any apparent actual occupational internal doses greater than 50 millirem committed effective dose equivalent (CEDE). The review also included the adequacy of evaluation of selected dose assessments, as appropriate, and included selected review of the program for evaluation of potential intakes associated with hard-to-detect radionuclides (e.g., transuranics). The inspectors selectively reviewed in-plant source term evaluations including average energy determinations. The inspectors reviewed airborne radioactivity control and monitoring for job coverage and selectively reviewed use of continuous air monitors.

During the inspection, the inspectors also reviewed: the adequacy and effectiveness of routine contamination control and monitoring practices; evaluated the adequacy of contamination detection capabilities; evaluated the extent of station contamination, and evaluated the frequencies and magnitude of personnel contamination events; and evaluated the detection of contamination beyond established barriers for the radiological controlled area (RCA). In addition, the inspectors also evaluated and reviewed the radiation dose consequences of the personnel contaminations. The inspectors evaluated the frequencies and magnitude of internal contaminations of personnel.

The inspectors reviewed and discussed high radiation area controls including high-dose rate and very high radiation area controls with radiation protection supervisors and technicians to identify changes that could potentially reduce program effectiveness and level of worker protection. The inspectors reviewed high radiation area access controls to the Unit 2 torus. The inspectors observed and conducted a selective review of high radiation area program procedures with radiation protection supervisors.

.3 Radiation Worker and Radiation Protection Technician Proficiency

During station tours, the inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors selectively questioned workers to determine if they were aware of the radiological conditions in their workplace; their RWP controls/limits in place; and that their performance took into consideration the level of radiological hazards present.

The inspectors observed radiation protection technician performance with respect to radiation protection work requirements to determine if they were aware of the radiological conditions in their workplace and the RWP controls/limits, and if their performance was consistent with expectations for potential radiological hazards present.

The inspectors reviewed radiological problem reports since the last inspection to identify radiation worker or radiation protection errors traceable to a similar cause. Corrective actions were reviewed, as appropriate.

.4 Problem Identification and Resolution

The inspectors selectively reviewed self-assessments and audits, as applicable, since the previous inspection to determine if identified problems were entered into the CAP for resolution. The inspectors evaluated the database for repetitive deficiencies or significant individual deficiencies to determine if self-assessment activities were identifying and addressing the deficiencies. The review also included evaluation of data to determine if any problems involved performance indicator (PI) events with dose rates greater that 25 R/hr at 30 centimeters, greater than 500 R/hr at 1 meter or unintended exposures greater than 100 millirem total effective dose equivalent (TEDE), 5 rem shallow dose equivalent (SDE), or 1.5 rem lens dose equivalent (LDE).

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02 - 14 Samples)

a. Inspection Scope

The inspectors conducted the following activities to determine if Exelon was properly implementing operational, engineering, and administrative controls to maintain personnel exposure as low as is reasonably achievable (ALARA). Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, applicable industry standards, and applicable station procedures. The inspectors also conducted reviews of Exelons outage pre-planning activities and implementation of those plans for the Unit 2 refueling and maintenance outage.

.1 Inspection Planning

The inspectors reviewed pertinent information regarding station collective dose history, current exposure trends, and ongoing or planned activities in order to assess current performance and exposure challenges. The inspectors determined the plant=s current 3-year rolling average collective exposure and determined the site specific trends in collective exposures (using NUREG-0713) and plant historical data.

The inspectors reviewed Unit 2 refueling and maintenance planned outage work activities. For this review, the inspectors selected work activities likely to result in the highest personnel collective exposures and reviewed the planning and preparation for those work activities to determine if ALARA requirements were integrated into work procedure and radiation work permit documents. The work activities selected included, but were not limited to: torus inspection (diving activities), under vessel work/control rod drive change-out, ISI, scaffolding activities, refueling activities, recirculation pump work, and valve work activities. The inspectors selectively reviewed implementation of lessons learned and operational experience. The inspectors evaluated adequacy of work time estimates for conduct of the work, versus that used for ALARA planning efforts. The inspectors evaluated shielding efforts as compared to shielding packages requested.

The inspectors evaluated use of benefits of water filled components to provide shielding, as applicable.

During the Unit 2 outage, the inspectors reviewed on-going and completed work activities to identify the adequacy and effectiveness of planning efforts to reduce radiation exposures ALARA. The inspectors evaluated interfaces between operations, radiation protection, maintenance planning, scheduling, and other groups for interface problems or missing program elements. The inspectors reviewed department ALARA plans for exposure reduction efforts. The inspectors reviewed occupational exposure performance associated with those activities that presented higher radiological risk potential. These tasks included control rod drive replacement, refueling work activities, in-vessel inspections, reactor vessel disassembly, main steam line valve work, and turbine and condenser work activities, including moisture separator work, in-service inspection, and torus diving activities. The inspectors toured the radiological controlled areas, including the Unit 2 drywell, and torus, and observed efforts to minimize occupational radiation exposure.

.2 Verification of Dose Estimates and Exposure Tracking Systems

The inspectors reviewed the assumptions and basis for current annual collective exposure estimates. The inspectors reviewed the exposure tracking system to evaluate the level of detail, and exposure report timeliness. The inspectors reviewed the methods used for adjusting exposure estimates, or replanning work when unexpected changes in scope or emergent work are encountered. The inspectors selectively reviewed contingencies implemented for work exhibiting elevated dose rates.

The inspectors selectively reviewed exposure results achieved, for the above tasks, with the intended dose established in ALARA plans for the work activities. The inspectors reviewed post-job evaluations and bases for additional exposures sustained for selected work activities.

.3 Source-Term Reduction and Control

The inspectors reviewed and discussed Exelon=s understanding of the plant source-term, including knowledge of input mechanisms to reduce the source term; and the source-term control strategy in place. The inspectors evaluated Exelon=s efforts to reduce radiation exposure including modified reactor shutdown and reactor coolant clean-up practices. The inspectors reviewed contingency plans for potential changes in source term and changes in plant source term as well as implementation of lessons learned. The inspectors reviewed source term controls and radiation exposure mitigation for reactor cavity drain-down, including implementation of lessons learned.

The inspectors discussed licensee reviews of Unit 2 chemistry controls for shut-down and discussed source term management and levels during the Unit 2 outage. The inspectors also evaluated implementation of Exelons procedurally described program for source term control, including its Five Year Exposure Reduction Plan. The inspectors also reviewed implementation of lessons, in the area of source term control and occupational exposure control, from previous outages.

During the Unit 2 outage, the inspectors reviewed performance relative to planned source term reduction initiatives.

.4 Radiation Worker and Radiation Protection Technician Performance

The inspectors observed radiation worker and radiation protection technician performance to determine if workers demonstrated exposure reduction practices. The inspectors also reviewed radiation protection technician performance to determine whether training/skill level was sufficient with respect to radiological hazards and the work involved.

.5 Declared Pregnant Workers

The inspectors selectively reviewed exposure results and exposure controls for declared pregnant workers with respect to exposure limitation requirements of 10 CFR 20.

.6 Problem Identification and Resolutions

The inspectors selectively reviewed applicable self-assessments, audits, and special reports related to the ALARA program since the last inspection. The inspectors evaluated if identified problems were entered into the CAP. (See Section 4OA2.4)

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03 - 6 Samples)

a. Inspection Scope

The inspectors reviewed selected activities, and associated documentation, in the below listed areas. The evaluation of Exelon=s performance in these areas was against criteria contained in 10 CFR 20, applicable TSs, and applicable station procedures.

Additionally, during the Unit 2 outage the inspectors toured the station, observed on-going work activities and evaluated instrument use to determine if appropriate radiological instrumentation was in use and had been properly source checked for operability prior to use.

.1 Calibration, Operability, Alarm Setpoint

The inspectors selectively reviewed calibration of various radiation safety instrumentation for various on-going work activities, including: Unit 2 torus diving, control rod drive work and refueling floor activities. (RO2 -332859, 79494; AMP-100-76118; SAC-4- 76667; telepole- 79471, 79163, 75616; RM-14- 74079, 74041; Lapel Air sampler - 78654, 79454, 78574). The inspectors also reviewed dosimetry for divers (Dive Pack 10318, 10315, 10308, 10301, 10320). The inspectors selectively reviewed operability checks; calibration, including use of appropriate sources; and alarm set-points, as applicable.

.2 Problem Identification and Resolution

The inspectors reviewed audits and self-assessments in this area to determine if identified issues in this area were entered into the CAP. The inspectors reviewed CRs and action requests (ARs) to evaluate Exelon=s threshold for identifying, evaluating, and resolving problems in this area. (See Section 4OA2.4)

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety

2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (71122.01 - 1 Sample)

.1 Onsite Inspection

a. Inspection Scope

The inspectors selectively verified that the licensee was continuing to implement the voluntary Nuclear Energy Institute (NEI)/Industry Ground Water Protection Initiative.

The inspectors selectively reviewed changes, monitoring results, leakage or spill events, and entries made into 10 CFR 50.75(g) records, remediation actions taken, and voluntary reporting of leaks and spills.

b. Findings

No findings of significance were identified.

2PS2 Radioactive Material Processing and Transportation (71122.02 - 4 Samples)

.1 Inspection Planning/In-Office Inspection (1 Sample)

a. Inspection Scope

The inspectors reviewed the solid waste system description in the UFSAR and recent radiological effluent release reports for information on the types and amounts of radioactive waste. The inspectors reviewed Exelons audit program in this area. (See Section 4OA2)

b. Findings

No findings of significance were identified.

.2 System Walkdown (1 Sample)

a. Inspection Scope

The inspectors toured the facility and visually inspected, where assessable, the station's radioactive liquid and solid waste collection, processing, and storage systems and locations to: determine if systems and facilities were consistent with descriptions provided in the UFSAR; evaluate materials conditions; and to identify changes made to systems. The inspectors also discussed operation of the systems with licensee personnel. The inspectors reviewed the following matters:

• The status of any non-operational or abandoned radioactive waste process equipment and the adequacy of administrative and physical controls for those systems;

• Changes made to radioactive waste processing systems and potential radiological impact including conduct of safety evaluations of the changes, as necessary;

• Current processes for transferring radioactive waste resin and sludge to shipping containers and mixing and sampling of the waste, as appropriate;

• Radioactive waste and material storage and handling practices;

• Sources of radioactive waste at the station (waste streams), processing (as appropriate) and handling of the waste; and

• Material conditions.

The review was against criteria contained in the station=s UFSAR, 10 CFR Part 20, 10 CFR 61, the Process Control Program (PCP), and applicable station procedures.

b. Findings

No findings of significance were identified.

.3 Waste Characterization and Classification (1 Sample)

a. Inspection Scope

The inspector reviewed and discussed the following matters:

• Radio-chemical sample analysis results for radioactive waste streams;

• The development of scaling factors for difficult to detect and measure radionuclides;

• Methods and practices to detect changes in waste streams;

• Classification and characterization of waste relative to 10 CFR 61.55 and 10 CFR 61.56;

• Implementation of applicable NRC Branch Technical Positions (BTPs) on waste classification, concentration averaging, waste stream determination, and sampling frequency;

• Current waste streams and their processing relative to descriptions contained in the UFSAR and the station=s approved PCP;

• Current processes for transferring radioactive waste resin and sludge discharges into shipping/disposal containers to determine adequacy of sampling; and

• Revisions of the PCP and the UFSAR to reflect changes (as appropriate).

The review was against criteria contained in 10 CFR 20, 10 CFR 61, 10 CFR 71, the UFSAR, the PCP, applicable NRC BTPs, and licensee procedures.

b. Findings

Introduction:

An NRC-identified Green NCV of 10 CFR 20, Appendix G, Section III.C. 5.

was identified due to Exelon not conducting a Quality Assurance Program sufficient to assure conformance with 10 CFR 61. Specifically, Exelons 10 CFR 61 Quality Assurance Program was not of sufficient depth to identify incorrect gamma spectroscopy analyses of a principal gamma emitting radionuclide used to scale hard-to detect radionuclides for purposes of waste classification in accordance with 10 CFR 61.55.

Description:

In July 2008, Exelon packaged and shipped a container (liner) of used resins for disposal at a licensed disposal facility. The used resins (Shipment No. PW-08-23) were originally loaded into the container (liner) in August 2006. Exelon had maintained the liner in controlled storage from the date of loading until shipment in July 2008. For purposes of conformance and classification of the shipment relative to 10 CFR 61, Exelon classified the contents of the liner, as a Class A shipment. This classification was based on a waste stream source term derived from samples of resins collected throughout 2005 (Reference: 10 CFR 61 Report dated July 7, 2006).

The inspector reviewed the technical bases for classifying the shipment as a Class A shipment, and noted that the station had experienced changes in its reactor coolant source term at both Units 2 and 3 due to identified fuel leakage. Exelons tracked fuel PIs showed an increase throughout 2005 and a leveling off in 2006. The inspector identified that, based on the changes in fuel performance, and that the loading of the waste liner had occurred in August 2006, the 2006 source term, appeared more appropriate for use in classifying the shipment for disposal. The 2006, 10 CFR 61 source term was derived from similar resin samples collected throughout 2006.

(Reference: 10 CFR 61 Report dated September 26, 2007). The 10 CFR 61 source term data for 2006, which included revised scaling factors, was available in September 2007, approximately ten months before shipment of the liner in July 2008.

Exelon used Cs-137 as a scaling nuclide to quantify Sr-90 for purposes of waste classification. The 2006 source term data showed an increase in the Sr-90 to Cs-137 ratio (scaling factor) indicating apparent increases in Sr-90 in the source term. Exelon subsequently re-analyzed the shipment classification, using the 2006 source term and associated 10 CFR 61 data and scaling factors. The re-analysis indicated an apparent change in the classification (Class B), using the 2006 source data, rather than the original classification (Class A) based on the 2005 source term data. Exelon immediately suspended shipments associated with this waste stream, placed this matter into its CAP, and initiated an EOC review.

As part of its review, Exelon requested its vendor to re-validate the sample analyses for the 2006 source term analysis used for development of 10 CFR 61 scaling factors presented in its September 26, 2007, 10 CFR 61 report. The vendor subsequently informed Exelon that its analysis of Cs-137 for 2006 samples was in error. The vendor had used an incorrect geometry during sample analysis which caused a lower Cs-137 indicated concentration. Consequently, this lower concentration gave a false indication of an increase in the Sr-90/Cs-137 scaling factor and, thus, a more restrictive waste classification (i.e., Class B).

Exelon did not conduct a Quality Assurance Program sufficient to assure conformance with 10 CFR 61. Specifically, the Quality Assurance Program was not sufficiently robust to identify errors in laboratory sample analysis results or changes in source term on a real time basis that could impact waste classification. Although Exelon had identified a change in the scaling factor in its analysis of 2006 10 CFR 61 source term data, it did not thoroughly evaluate the analysis results sufficient to detect errors in laboratory analysis that were subsequently revealed through NRC inspector questioning. In addition, the inspector identified that Exelon had identified, in a May 2008 audit, that it had not developed a tracking index to identify changes in scaling factors as required by procedures. However, the audit did not identify laboratory analysis concerns.

The failure to conduct a sufficiently robust 10 CFR 61 Quality Assurance Program, to assure conformance with 10 CFR 61, is a performance deficiency that was reasonably within Exelons ability to foresee and correct, and which should have been prevented.

Although Exelon had identified in its May 14, 2008, audit that no trending index existed to identify potential shifts in 10 CFR 61 scaling factors for hard to identify radionuclides that would require additional or more frequent sampling, subsequent Exelon corrective action analyses did not identify an existing error in a 10 CFR 61 laboratory sample analysis result reflected in its September 26, 2007, 10 CFR 61 scaling factor report.

Analysis:

This issue is not subject to traditional enforcement in that it did not have actual safety consequence, it was not an issue that had the potential to impact the NRCs ability to perform its regulatory function, and there were no willful aspects.

The finding is more than minor in that: 1) the finding is associated with the NRC Manual Chapter 0609 Public Radiation Safety Cornerstone attribute of Program and Process Transportation Program, and 2) the finding did affect the associated cornerstone objective to ensure adequate protection of public health and safety. Specifically, Exelons 10 CFR 61 waste classification program did not identify incorrectly analyzed waste samples used to classify radioactive waste shipments, in accordance with 10 CFR 61 for land disposal.

Exelon re-calculated the classification of the shipment, using the corrected laboratory analysis results from its 2007 Part 61 Report, which confirmed that the original classification, made with data from its 2006 10 CFR Part 61 report was acceptable and indicated the shipment remained Class A.

This finding in the area of 10 CFR 61 was evaluated against criteria specified in NRC Manual Chapter 0609, Appendix D, and determined to be of very low safety significance (Green). Specifically, no radiation limits were exceeded, there was no breach of packaging, there was no certificate of compliance finding, there was no low level burial ground non-conformance, and lastly, there was no failure to make notifications or provide emergency notification information. The inspectors determined that the cause of this finding was related to the cross-cutting area of PI&R, of self and independent assessments component, in that, although actions were taken to coordinate and communicate results from assessments to affected personnel, corrective actions were not sufficiently comprehensive to identify incorrect vendor analyses. (P.3(c))

Enforcement:

Title 10 of the Code of Federal Regulations (CFR), Part 61, requires, in part, a Quality Assurance program sufficient to assure compliance with 10 CFR 20, Appendix G, Section III.C.5. Contrary to the above, prior to July 2008, Exelons 10 CFR 61 quality assurance program was not adequate to identify incorrect gamma spectroscopy analyses of a principal gamma emitting radionuclide used to scale hard-to detect radionuclides for purposes of waste classification in accordance with 10 CFR 61.

This is an NRC-identified violation of 10 CFR 20, Appendix G, Section III.C. 5. Because this matter was of very low safety significance, and has been entered into the CAP (IR 799894), this violation is being treated as a NCV consistent with Section VI.A

.1 of the

NRC Enforcement Policy: NCV 05000277; 278/2008004-02, Failure to Comply with 10 CFR 20 Appendix G.

.4 Shipment Preparation

a. Inspection Scope

The inspectors selectively reviewed shipment preparation and the training and qualification program for personnel handling, packaging, and shipping radioactive materials during NRC Integrated Inspection 05000277/2008002 and 05000278/2008002 (Reference ADAMS ML081270699).

b. Findings

No findings of significance were identified.

.5 Shipment Records and Documentation (1 Sample)

a. Inspection Scope

The inspectors selected and reviewed the records associated with five non-excepted shipments of radioactive material made since the previous inspection in this area (Shipment Nos. PW-08-08-10, PW-07-029, PW-08-11, PW-08-21, and PW-08-23).

The following aspects of the radioactive waste, radioactive material packaging, and radioactive material shipping activities were reviewed.

• Implementation of applicable shipping requirements including completion of waste manifests;

• Implementation of the specifications in applicable Certificates of Compliance, as appropriate, for the approved shipping casks including limits on package contents;

• Classification and characterization of waste relative to 10 CFR 61.55 and 61.56, as appropriate;

• Implementation of recent NRC and Department of Transportation (DOT) shipping requirements rule changes;

• Implementation of 10 CFR 20 Appendix G;

• Implementation of specific radioactive material shipping requirements;

• Packaging of shipments;

• Labeling of shipping containers;

• Placarding of transport vehicles;

• Conduct of vehicle checks;

• Provision of driver emergency instructions;

• Completion of shipping paper/disposal manifest;

• Evaluation of package against package performance standards, as appropriate;

• Conformance with procedures for cask loading, closure and use requirements including consistency with cask vendor approved procedures; and

• Use of latest revision documents.

The review was against criteria contained in 10 CFR 20; 10 CFR 61; 10 CFR 71; applicable DOT requirements, as contained in 49 CFR 170-189 for the above areas; station procedures; applicable disposal facility licenses; and applicable Certificates of Compliance or vendor procedures for various shipping casks.

The inspectors also selectively reviewed the year 2007 Annual Radioactive Effluent Release Report, relative to types and quantities of radioactive waste shipped offsite and relative to changes to the PCP.

b. Findings

No findings of significance were identified.

4OA1 PI Verification (71151 - 4 Samples)

.1 Barrier Integrity PIs

a. Inspection Scope

The inspectors reviewed a selected sample of PBAPSs information submitted for the four Barrier Integrity PIs listed below to assess the accuracy and completeness of the data reported to the NRC for these PIs. The PI definitions and the guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 5, and licensee procedure LS-AA-2001, Collecting and Reporting of NRC PI Data, were used to verify that the reporting requirements were met. The inspectors reviewed a selected sample of the raw PI data collected since July 2007 to July 2008 and compared graphical representations from the most recent PI report to the raw data to verify the data was included in the report. The inspectors discussed the methods for compiling and reporting the PIs with the cognizant chemistry and engineering personnel. The inspectors also examined a selected sample of surveillance test procedures and records to verify the PI data was appropriately captured for inclusion into the PI report and the individual PIs were correctly calculated.

• Unit 2 and Unit 3 Reactor Coolant System (RCS) Specific Activity; and

• Unit 2 and Unit 3 RCS Leakage.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152 - 1 Annual Sample)

.1 Annual Sample - Breaker Coordination (1 Sample)

a. Inspection Scope

The inspectors reviewed Exelons evaluation and corrective actions associated with several breaker coordination issues. The inspectors reviewed CRs and the associated actions against the requirements of Exelons CAP to ensure that the full extent of the issues were identified, appropriate evaluations were performed, and appropriate corrective actions were specified and prioritized. The inspectors interviewed relevant station personnel and reviewed applicable station procedures to ensure that the issues were addressed appropriately.

b. Findings and Observations

No findings of significance were identified. The inspectors determined that Exelons proposed corrective actions were reasonable with respect to the breaker coordination issues. Exelon performed an appropriate EOC and implemented calculation updates that reflected the current plant configuration. The inspectors determined that Exelon followed the appropriate configuration change review procedures to determine that the coordination issues do not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. The inspectors agreed with Exelons conclusion that safe shutdown is not adversely affected by the coordination issues.

.2 Routine Review of Items Entered into the CAP

a. Inspection Scope

As required by IP 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures and human performance issues for follow-up, the inspectors performed routine screening of issues entered into PBAPSs CAP. The review was accomplished by selectively reviewing copies of Issue Reports (IRs) and accessing PBAPSs computerized database.

b. Findings

No findings of significance were identified.

.3 ISI Examinations

a. Inspection Scope

The inspectors reviewed a sample of examination reports and CRs initiated during ISI examinations to evaluate the licensee=s effectiveness in the identification and resolution of problems. The CRs reviewed are listed in Attachment A-1. The inspectors verified that the nonconforming conditions identified were reported, characterized, and entered into the CAP.

b. Findings

No findings of significance were identified.

.4 Review Radiation Safety IRs (71121.01, 71121.02, 71121.03, 712202, 71151)

a. Inspection Scope

The inspectors reviewed ARs to evaluate Exelon=s threshold for identifying, evaluating, and resolving problems, including identifying and implementing effective corrective actions. The review included a check of possible repetitive issues such as radiation worker or radiation protection technician errors. The documents reviewed are listed in the Attachment.

This review was against the criteria contained in 10 CFR 20, TSs, and the station procedures. One finding that was associated with the cross-cutting area of PI&R is discussed in Section 2PS2.3.

b. Findings

No findings of significance were identified.

4OA3 Event Followup (71153 - 2 Samples)

.1 Number 1 Autotransformer Failure and Fire

a. Inspection Scope

The inspectors reviewed the issues associated with the Number 1 autotransformer fire event that occurred in the north switchyard on July 23 and 24, 2008. The inspectors discussed the event with licensee management, engineering, operations, and electrical distribution maintenance personnel to gain an understanding of the conditions leading up to the fire and actions taken following the event to assess PBAPSs actions. The inspectors also reviewed the event for reportability in accordance with NUREG 1022, Event Reporting Guidelines, Revision 2.

b. Findings

Introduction:

The inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(2)(xi) because the NRC Operations Center was not notified of a reportable event. Specifically, PBAPS did not formally report, to the NRC Operations Center, a planned press release and the notification of other government agencies regarding a transformer fire and petroleum product spill event that occurred on July 23 and 24, 2008.

Description:

At about 11:25 p.m., on July 23, 2008, an electrical transient caused one of the two, TS required off-site power sources to the Peach Bottom Stations 4kV emergency busses to isolate. The impacted 4kV emergency busses for both units successfully fast-transferred to the alternate source. The momentary loss of power experienced during the transfer resulted in several isolations including reactor water cleanup and hydrogen water chemistry trips on both units. The transient also resulted in loss of an alternate off-site power source that was aligned to feed station auxiliary loads at the time of the event. Isolation of the alternate off-site source resulted in loss of power to various administrative buildings and other equipment.

Coincidently, the main control room received a report of a fire located at the North Substation. Plant equipment operators responded to the North Substation and found a fire on the 'A' phase of the autotransformer connecting the 500kV and 230kV transmission systems. Analysis of the event showed that the transformer experienced a phase-to-ground fault resulting in a transformer tank rupture, an oil fire, and autotransformer failure. The apparent cause of the failure was associated with a high voltage bushing. Both units remained stable, but the station entered a seven-day TS action statement as a result of the loss of one of the two required off-site power sources.

The second off-site power source was restored approximately seven hours following the event allowing the station to exit the TS action statement.

During the transformer fire, oil migrated outside the containment dike resulting in a ground fire. The station fire brigade responded. The local government emergency dispatcher was contacted to request assistance and the local fire company responded.

The ground fire was extinguished and the transformer fire controlled. Hazardous material personnel responded to the breach in the transformer containment. The event resulted in ground remediation due to the oil migration. On July 24, 2008, offsite organizations, including the Pennsylvania Department of Environmental Protection (PA DEP) were notified of a spill greater than 50 gallons and the environmental aspects of the fire. The inspectors noted that special event procedure (SE) -6, Pollution Incident Protection Procedure, Attachment 1, Petroleum Product Spill External Notification Flow Chart, requires the reporting of a release of greater than 25 gallons of petroleum product to the PA DEP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also, on July 24, 2008, Exelon issued a press release for the fire event.

The inspectors determined that the failure to formally notify the NRC regarding a planned press release and notification of State and local government agencies for the occurrence of an event related to the health and safety of the public and protection of the environment as required by 10 CFR 50.72 was a performance deficiency. This event was related to pubic health and safety because it involved a fire emergency that contributed to the loss of two of plants three offsite power sources. This event was related to protection of the environment because it involved the spill of greater than 50 gallons of oil.

Analysis:

This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. While reviewing this finding, the inspectors considered the fact that the NRC was informally notified. The inspectors also noted that subsequent to this event PBAPS initiated IRs (811186, 815301, 800977 and 799684) in the CAP to review enhancement of tools used to decide when reporting to the NRC is required for the issuance of press releases and notifications of other government agencies. The inspectors considered the above and evaluated the severity of this violation using the criteria contained in Supplement I - Reactor Operations and Section VI.A.1 of the NRCs Enforcement Policy and determined that this finding met the criteria for disposition as a NCV.

Enforcement:

Title 10 of the Code of Federal Regulations (CFR) Part 50.72(b)(2)(xi)states, in part, Four-hour reports - the licensee shall notify the NRC as soon as practical and in all cases, within four hours of the occurrence of any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Contrary to these requirements, on July 24, 2008, Exelon failed to make the appropriate four-hour report when an event related to the health and safety of the public and protection of the environment occurred, and a press release and notification of other government agencies was made. Specifically, on July 24, 2008, offsite organizations, including the Pennsylvania Department of Environmental Protection (PA DEP) were notified of a spill greater than 50 gallons and the environmental aspects of the fire. Because the licensee has entered this issue into their CAP (IRs 811186, 815301, 800977 and 799684), the issue is being treated as a Severity Level IV NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000277; 278/2008004-03, Failure to Make a 10 CFR 50.72(b)(2)(xi)

Notification.

.2 Multiple ESW System Through-wall Leaks

a. Inspection Scope

The inspectors reviewed PBAPSs actions taken in response to an event that began on July 6, 2008, when a through-wall leak was identified at a piping elbow in the ESW system that provides cooling water to the DG heat exchangers. The inspectors discussed the event with licensee management, engineering, NDE, and maintenance personnel to gain an understanding of the conditions leading up to the leak identification and the actions taken immediately following to assess licensee actions.

Since this component was an elbow, GL 90-05 was used for guidance to determine operability. Per GL 90-05 requirements, five augmented examinations were scheduled, within 15 days, on similar locations. The augmented examinations identified another elbow leak. Also, a leak was visually observed in a valve body. Because valves are not components within the scope of GL 90-05, the valve body leak resulted in declaring the component inoperable and the valve was immediately replaced. Visual inspections were performed on piping in all the DG rooms. The visual inspections found four additional piping leaks. All of the remaining leaks were evaluated per GL 90-05 or ASME Code Case 513 based on the component type. Based on the failure analysis results of selected replaced components, under deposit pitting corrosion that may be influenced by biological activity was preliminarily determined to be the cause of the degradation. The inspectors also reviewed PBAPS plans to conduct further failure analysis and evaluation.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status reviews and inspection activities.

b. Findings

No findings of significance were identified.

.2 Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review

a. Inspection Scope

The inspectors reviewed the final report for the INPO plant assessment of PBAPS conducted in January 2008. The inspectors reviewed the report to ensure that issues identified were consistent with the NRC perspectives of licensee performance and to determine if any significant safety issues were identified that required further NRC follow-up.

b. Findings

No findings of significance were identified.

.3 Independent Spent Fuel Storage Installation (ISFSI) Radiological Controls (60855.1)

a. Inspection Scope

The inspectors toured the ISFSI facility and conducted independent gamma radiation surveys of the ISFSI facility and compared the results to previous surveys. The inspectors also observed and evaluated implementation of radiological controls, including RWPs and postings, and discussed the controls with radiation protection personnel. The inspectors reviewed radiation surveys as well as environmental radiation monitoring data from thermoluminescent dosimeters around the facility. The inspectors reviewed implementation of applicable licensee conditions associated with module temperature control.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On October 17, 2008, the resident inspectors presented the inspection results to Mr. W. Maguire and other PBAPS staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations

None.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Exelon Generation Company Personnel

W. Maguire, Site Vice President

G. Stathes, Plant Manager

J. Armstrong, Regulatory Assurance Manager

C. Behrend, Engineering Director

L. Bunner, Work Management Director

L. Lucas, Chemistry Manager

R. Franssen, Operations Director

R. Holmes, Radiation Protection Manager

D. DeBoer, Security Manager

T. Wasong, Training Director

NRC Personnel

F. Bower, Senior Resident Inspector

M. Brown, Resident Inspector

J. DAntonio, Sr. Operations Engineer

D. Kern, SRI, TMI

R. Nimitz, Senior Health Physicist

T. OHara, Reactor Inspector

J. Tifft, Reactor Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None.

Opened/Closed

05000277; 278/2008004-01 NCV Inadequate EOC Review Results in Delay in Discovery of ESW Leaks (Section 1R15)

05000277; 278/2008004-02 NCV Failure to Comply with 10 CFR 20 Appendix G (Section 2PS2.3)

05000277; 278/2008004-03 NCV Failure to Make a 10 CFR 50.72(b)(2)(xi)

Notification (Section 4OA3.1)

Closed

None.

Discussed

None.