TMI-12-047, (TMI-1), Biennial 10 CFR 50.59 and Commitment Revision Reports for 2010 and 2011

(TMI-1), Biennial 10 CFR 50.59 and Commitment Revision Reports for 2010 and 2011

ML12109A235
Person / Time
Site:	Three Mile Island
Issue date:	04/16/2012
From:	Newcomer M Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, NRC Region 1
References
TMI-12-047
Download: ML12109A235 (19)

Text

Exelon.

Three Mile Island Unit 1 Telephone 717-948-8000 Nuclear Route 441 South, P.O. Box 480 Middletown, PA 17057 April 16, 2012 TMI-12-047 U. S. Nuclear Regulatory Commission, Region I Attn: Document Control Desk Washington, D.C. 20555 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI-1)

RENEWED OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289

SUBJECT:

BIENNIAL 10 CFR 50.59 AND COMMITMENT REVISION REPORTS FOR 2010 AND 2011 Enclosed are the 2010-2011 Biennial 10 CFR 50.59 and Commitment Revision Reports as required by 10 CFR 50.59 (d)(2) and SECY-00-0045 (NEI 99-04).

There are no regulatory commitments contained in this transmittal.

If you have any questions or require additional information, please contact David Atherholt, of Regulatory Assurance, at 717-948-8364.

Sincerely, Mark Newcomer Plant Manager, TMI-1 Exelon Generation Co., LLC MN/mdf Enclosure cc: USNRC, Regional Administrator, Region I USNRC, Senior Resident Inspector, TMI Unit 1

Region I Administrator, cc: USNRC RegionalManager, TMI TMI USNRC Project Resident Inspector, USNRC Senior

2010-2011 Biennial IOCFR 50.59 and Commitment Revision Reports THREE MILE ISLAND UNIT 1 DOCKET NO. 50-289 BIENNIAL 10CFR 50.59 AND COMMITMENT REVISION REPORTS TABLE OF CONTENTS 10CFR 50.59 Report 2 Modifications 3 Procedure Changes 9 Commitment Revision Report 10 1

2010-2011 Biennial IOCFR 50.59 and Commitment Revision Reports EXELON CORPORATION THREE MILE ISLAND UNIT 1 DOCKET NO. 50-289 BIENNIAL 10CFR 50.59 REPORT JANUARY 1, 2010 THROUGH DECEMBER 31, 2011 10CFR50.59 EVALUATION SUMMARIES 2

2010-2011 Biennial IOCFR 50.59 and Commitment Revision Reports Modifications

Title:

TMI-1 Once Through Steam Generator Replacement Year Implemented: 2010 Evaluation Number: Engineering Change Request No. TM 06-00935/Safety Evaluation No.SE-000224-030 Rev 0 Brief

Description:

SE-000224-030 Rev 0 was approved by PORC on 10/3/08. Since that time AREVA has completed the documents that were carried as Unverified Assumptions (UVAs) in the original 50.59 evaluation, the Steam Generator Replacement Project Engineering FASA recommended some changes to the 1 OCFR50.59 for ECR TM 06-00935 including the 10CFR50.59 evaluation and the NRC has requested that we address a Large Break Loss-of Coolant Accident in the upper hot leg.

ECR 06-00935 documents the technical evaluation and qualification of the replacement steam generators being installed at Three Mile Island Unit 1 (TMI-1). The original Once Through Steam Generators (OTSGs), supplied by Babcock & Wilcox, will be replaced with new Enhanced Once Through Steam Generators (EOTSGs) manufactured by AREVA NP Inc. (AREVA NP). This 10CFR 50.59 (50.59) review evaluates the impact on the TMI-1 Updated Final Safety Analysis Report (UFSAR) to ensure the EOTSGs can be installed under the current plant license.

The shell, the outside of the tubes, and secondary side of the tubesheets form the boundaries of the steam-producing volume of the vessel (the secondary side). The shell of the EOTSG is fabricated from high strength SA 508 Gr. 3 Cl. 2 forgings, welded circumferentially. This replaces the lower strength SA 212-B shell used for the OTSGs which included longitudinal welds, eliminated in the EOTSG design, as well as circumferential welds. The higher strength material allows the EOTSG shell to be thinner resulting in a larger secondarX inventory for an equivalent outer volume. The net increase in secondary volume is 250.61 ft , or approximately 7.34%.

Due to this change in Secondary volume there are minor adverse effects that were subjected to a 50.59 Evaluation. The Evaluation demonstrates that the impact on accidents affected by the secondary side volume increase is not significant and will not lead to consequences beyond those in the UFSAR.

In addition to the increase in secondary volume existing EFW models for the OTSG considered carbon steel Tube Support Plates (TSPs) and shroud; thus thermal expansion was not a concern and the cold gap was maintained at operating temperature. Differential thermal expansion between the Martensitic stainless steel TSPs in the EOTSGs and the carbon steel shroud results in an increased gap around the periphery of the 1 5 th TSP as the plant heats from ambient to operating temperature. This gap will marginally increase EFW bypass to the EOTSG tubes and, therefore, adversely affects the ability of EFW to perform its accident mitigation function. This change was also subjected to a 50.59 Evaluation.

Changes to the Revision 0 10CFR50.59 evaluation include:

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2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports

1. Deletions of all UVAs as these have been cleared.

2. Addition of Large Break LOCA in the upper hot leg to questions 2 and 4 to address whether the probability/consequences of a steam generator tube failure increase due to a large break LOCA in this area.

3. Deletion of Large Break and Small Break LOCA discussions in Question 1, (Probability of Occurrence) as these have been determined to be non-adverse changes in the 50.59 screen,

4. Deletion of Large Break LOCA discussion in Question 3, (Increase in Consequences) as this has been determined to be a non-adverse change in the 50.59 screen,

5. Addition of Loss of Electric Load (LOEL) discussion in Question 7 (Affect on Design Basis Fission Product Barrier).

6. The response in Section 7.2.2 was revised to include a discussion similar to sections 3.6 and 4.1.2 for FWLB as SG Tube-to-shell delta T exceeded 60 degree F.

7. Changed the response to Question 8 (Change in Method of Evaluation) to address the use of RELAP5/MOD2-B&W for the Main Steam Line Break Evaluation and the EPRI endorsed methodologies for tube integrity. This discussion was originally in the 50.59 screen

8. Various editorial changes Summary of Evaluation:

As a result of the change in secondary volume and EFW bypass, several of the accident evaluations in Chapter 14 of the UFSAR were impacted. The 50.59 evaluation responds to the specific questions of 1 0CFR50.59 the summary provided below presents the principle arguments and is not meant to be exhaustive in all 10 CFR 50.59 Evaluation aspects. The reader is advised to refer directly to the 10 CFR 50.59 Evaluation for a complete treatment of these issues.

Loss of Coolant Accident (LOCA) - As described in the 50.59 Screen with the exception of the large break LOCA in the upper hot leg, the large break LOCA is unaffected. The large break LOCA in the upper hot leg applies a very large thermal load on the steam generator tubes and the tube-to-tubesheet welds. The loads associated with this temperature differential were evaluated to ensure the structural and leakage integrity criteria will continue to be met. The small break LOCA results in a maximum tube-to-shell delta T of -2040 F, shell is hotter than the tubes. The loads associated with this temperature differential were evaluated to ensure the structural and leakage integrity criteria will continue to be met. Additionally due to changes in EFW bypass of the steam generator the Peak Cladding Temperature (PCT) was evaluated to ensure the value in the UFSAR was not affected.

Steam Generator Tube Rupture (SGTR) - As discussed in the 50.59 Evaluation as a result of improvements in tube material and steam generator design the magnitude of degradation of the replacement steam generator tubes is demonstrated to be less. In addition loads associated with tube-to-shell differential temperatures were evaluated for several accidents that affect the steam generator and concluded that the acceptance criteria for tube-to-shell thermal loads would be met. Finally a SGTR was evaluated against the Analysis of Record (AOR) documented in the UFSAR and concluded that the AOR bounded the consequences of a replacement steam generator tube failure.

Main Steam Line Break (MSLB) - As described in the 50.59 Evaluation the dose consequence from a MSLB is less for the replacement steam generator as it is limited to a 1 gpm primary to secondary accident induced leak rate. The current steam generators due to existing 4

2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports degradation assume a higher leak rate. With respect to the post accident containment environment due to the increase in secondary volume pressure and temperature increase, pressure is bounded by post-LOCA pressures however temperature increases by -8o F. This increase was evaluated against the affect on environmentally qualified components, containment cooling and containment penetration pressurization and concluded that since the transient is of a shorter duration than the current analysis of record there is no impact to these issues. The transient is of a shorter duration due to the new Main Steam venturis, which are integral to the Main Steam lines at the exit points of the new steam generators.

Loss of Feedwater (LOFW) - As described in the 50.59 Evaluation the LOFW accident with off-site power available is the limiting event for EFW flow capability. The analysis performed to evaluate this accident demonstrated that the UFSAR acceptance criteria continue to be met.

Feedwater Line Break (FWLB) - As described in the 50.59 Evaluation the FWLB accident with offsite power available is similar to a LOFW except that flow is lost abruptly. Response to these events is similar with the exception that FWLB is specifically concerned with a SGTR due to axial compressive loads. Analysis concluded that these loads are less than those previously qualified in the ASME Code design analysis for the steam generators.

Loss of Electric Load (LOEL) - As described in the 50.59 Evaluation the LOEL this event can be looked at from three different scenarios 1) Loss of Load from Rated Power with Unit Runback,

2) Loss of Load with Reactor Trip and 3) Station Blackout. As these events result in lifting of the Main Steam Safety Valves however, the replacement steam generators do not affect the dose consequences of these events.

Changes in Methodology - The methodologies used for assessing Main Steam Line Break and Tube Integrity described in the UFSAR were changed to use other approved methodologies.

These discussions were originally contained in the 50.59 Screen but based on input from the FASA review a decision was made to move these discussions to the 50.59 Evaluation.

As described in the 50.59 Evaluation this evaluation concludes that prior NRC approval of the component is not required.

The analyses for some of the events impacting the replacement steam generators were performed at a tube plugging level of 5%. Thus if 5% tube plugging is approached additional analyses will need to be performed to justify higher levels of tube plugging.

Title:

Spent Fuel Pool Rerack Year Implemented: 2010 Evaluation Number: Engineering Change Request No. TM-08-00872/SE 08-00872 Brief

Description:

The 50.59 evaluation was performed for the Spent Fuel Pool Rerack Project, Phase 3, as described in ECR TM 08-00872. The evaluation was completed to document the replacement of the original neutron absorber (Boral) with an upgraded material (Metamic).

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2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports The material substitution is the only aspect of the project that requires a 50.59 evaluation since it affects the description in the original Safety Evaluation Report (SER) submittal for the rerack, which specified Boral as the neutron absorber material in the racks. The Phase 3 racks containing the Metamic material were installed in April 2009 per the referenced ECR. The ECR was approved under a 50.59 screening, which determined that the Metamic replacement was not an adverse change to any design description, function, or methodology previously evaluated for the rerack project. Subsequent discussions with Regulatory Affairs resulted in the decision to perform an evaluation based on the potential adverse change, due to the replacement of the absorber material as specified in the SER.

Summary of Evaluation:

The Metamic neutron absorber is functionally identical to Boral in that they use B4C as the absorber within a metallic matrix fabricated into a panel installed in the rack cell walls of the new Spent Fuel Pool racks. Metamic used in recent reracks projects in the industry is an improvement to Boral, as it maintains equivalent absorber properties while eliminating some imperfections (blisters, etc.) experienced with Boral. All interfaces and properties (fuel handling/storage, spent fuel pool cooling and chemistry, seismic design, criticality analyses) of Metamic have been evaluated as meeting or exceeding those of Boral.

Title:

Cycle 19 Core Reload Design Year Implemented: 2011 Evaluation Number: Engineering Change Request No. TM 10-00717 Brief

Description:

Proposed Change The change proposed by ECR TM 10-00717 is the core reload for TMI-1 Cycle 19 (T1C19) and the revised TMI-1 Core Operating Limits Report (COLR). The T1C19 reload uses Mark-B-HTP fuel designs previously reviewed and approved for use at TMI-1. Core operating limits were developed using NRC-approved AREVA methods contained in reload topical report BAW-10179P-A, Rev. 8. The BAW-10179 reload topical is identified in TMI-1 Technical Specification 6.9.5.2 as an analytical method that has been previously reviewed and approved by the NRC for use at TMI-1.

Difference Between Existing Requirements and the Proposed Change The T1C19 COLR was partially based on fuel thermal and mechanical analyses performed using AREVA's NRC-approved COPERNIC methodology (BAW-10231 P-A, Rev. 1) in place of TACO3 / GDTACO analyses performed for previous TMI-1 reloads. The TACO3 / GDTACO methodology is described in the TMI-1 UFSAR; the COPERNIC methodology is not. Both methodologies are used to determine design basis limits to prevent centerline fuel melt (CFM) and 1% cladding transient strain. Therefore, the T1C19 COLR does involve use of an alternative evaluation methodology that is used in establishing the design bases.

Affect on Nuclear Safety and Basis for that Determination 6

2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports The change from the TACO3 / GDTACO to COPERNIC methodology for fuel thermal and mechanical performance analyses for the T1 C19 reload does not have an adverse affect on nuclear safety. The COPERNIC code is AREVA's most current approved method for these analyses and eliminates the need for previously used adjustments to fuel rod thermal performance linear heat rate limits, as the TACO3 / GDTACO methodology did not explicitly treat the burnup-related degradation of fuel thermal conductivity.

The COPERNIC methodology has been approved by the NRC (BAW-1 0231 P-A) and has been approved by the NRC for application at B&W-1 77 plants via incorporation into AREVA's reload topical report BAW-10179P-A, Rev 8.

The implementation of COPERNIC for TI C19 does not reduce margins to operating limits.

T1 C19 reload analyses performed using COPERNIC-based CFM and 1% cladding transient strain limits justified the same Reactor Protection System axial power imbalance protective limits that were used in Cycle 18.

Regulatory Requirements that are Impacted and How any Discrepancies were Resolved There are no regulatory requirements impacted by this change. The COPERNIC methodology has been applied to the T1 C19 reload for fuel designs within the range of applicability approved for the methodology.

The COPERNIC methodology has been approved for use at TMI via incorporation into the AREVA reload topical report BAW-10179P-A, Rev. 8, which is identified in TMI-1 Technical Specification 6.9.5.2 as an analytical method that has been previously reviewed and approved by the NRC for use at TMI-1, with the current revision listed in the TMI COLR. The TiC19 COLR lists the current revision of BAW-10179P-A as Rev. 8.

10 CFR 50.59 Evaluation Conclusions The use of the COPERNIC code is appropriate for the intended application and within the limitations imposed by BAW-10231 P-A, Rev. 1. Therefore, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

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2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports Summary of Evaluation:

The change from the TACO3 / GDTACO to COPERNIC methodology for fuel thermal and mechanical performance analyses for the T1 C19 reload does not have an adverse affect on nuclear safety. The COPERNIC code is AREVA's most current approved method for these analyses and eliminates the need for previously used adjustments to fuel rod thermal performance linear heat rate limits, as the TACO3 / GDTACO methodology did not explicitly treat the burnup-related degradation of fuel thermal conductivity.

The implementation of COPERNIC for T1 C19 does not reduce margins to operating limits.

TI C19 reload analyses performed using COPERNIC-based CFM and 1% cladding transient strain limits justified the same Reactor Protection System axial power imbalance protective limits that were used in Cycle 18.

Title:

Alternative Design for OTSG Cold Leg Nozzle Dams Year Implemented: 2011 Evaluation Number: Engineering Change Request No. TM-11-00512 Brief

Description:

The current OTSG nozzle dam design including the attachment on the cold leg nozzle was established in ECR 06-00935 as part of the OTSG replacement modification. The alternative design described in ECR 11-00512 uses all of the structural elements as described in ECR 06-00935 except that a Scientech part no 90E0103 diaphragm seal (the seal used at Oconee) is substituted for the Scientech part no 90E1 056.

The alternative configuration is installed and operated the same as the existing approved seal.

The alternative configuration has been tested to demonstrate that leakage is low (< 2 GPM) with only the passive seal at the maximum system pressure. Test performed with pressure above 26 psig was completed successfully at TMI on 10/26/11. With the maximum possible level in the fuel transfer canal (346' 0" elevation), the water pressure at the bottom of the cold leg nozzle dams (286' 1.5" elevation) is less than 26 psig.

UFSAR Section 4.2.2.2 (Attachment 1) states that the passive seal will "limit the resulting leak rate to < 2 gpm at 30 psig water pressure."

The reduction in design margin (from 30 psig to 26 psig) for the passive seal was an adverse change to the UFSAR described function of the OTSG cold leg nozzle dam. Therefore, a safety evaluation was performed.

Summary of Evaluation:

The safety basis for use of the OTSG cold leg nozzle dams does not rely upon the performance of the active or passive seal performance. The structural integrity of the nozzle dam is sufficient to ensure there is no adverse affect on the dose consequence of any design basis accident. The safety analyses for applicable events (loss of inventory or loss of decay heat removal during refueling) are not affected. The evaluation concluded that this (or any) modification to the diaphragm seal does not require prior NRC approval.

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2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports Procedure Changes There were no 10CFR50.59 required procedure changes for this reporting period.

End of 1 OCFR50.59 Revision Report 9

2010-2011 Biennial I0CFR 50.59 and Commitment Revision Reports EXELON CORPORATION THREE MILE ISLAND UNIT 1 DOCKET NO. 50-289 BIENNIAL COMMITMENT REVISION REPORT JANUARY 1, 2010 THROUGH DECEMBER 31, 2011 10

2010-2011 Biennial 10CFR 50.59 and Commitment Revision Reports Letter Source: "GL 83-28 Required Actions Based on Generic Implications of Salem ATWS Events" Exelon Tracking No.: 1983T0131 Nature of Commitment: GPUN committed to the NRC to modify Technical Function (TF) procedure EP-017 (Corporate Procedure 5000-ADM-7370.02) to implement a program that will collect, review, and track all technical information received from vendors. Technical manual changes were excepted since they were handled through EP-021 (Corporate Procedure 5000-ADM-7316.03).

Summary of Justification:

Commitment Change Evaluation Form 10-05 revised this commitment from "Continuing" to "Historical". The referenced procedure EP-017 has been voided and replaced by CC-AA-204.

NRC GL 83-28 was superseded by GL 90-03 and GL 90-03 supplement 1. Commitment 1983T01 31 is superseded by commitments 1985T0060 and 1985T0062 which are within CC-AA-204.

Letter Source: "GL 83-28 Required Actions Based on Generic Implications of Salem ATWS Events" Exelon Tracking No.: 1985T0034 Nature of Commitment: GPUN committed to the NRC to implement the INPO-sponsored Nuclear Utility Task Action Committee (NUTAC) Vendor Equipment Technical Information Program (VETIP).

Summary of Justification:

Commitment Change Evaluation Form 10-06 revised this commitment from "Continuing" to "Historical". NRC GL 83-28 was superseded by GL 90-03 and GL 90-03 supplement 1.

Commitment 1985T0034 is superseded by commitments 1985T0060 and 1985T0062 which are within CC-AA-204.

Letter Source: 5211-85-2160 "GL 83-28 ATWS Vendor Interface Manual Document Control NUTAC" Exelon Tracking No.: 1985T0062 / 1122355-08 Nature of Commitment: GPUN committed to the NRC to review and revise as necessary GPUN procedures 5000-ADM-7316.02 and 5000-ADM-7370.02 for consistency with Nuclear Utility Task Action Committee (NUTAC) Vendor Equipment Technical Information Program (VETI P).

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2010-2011 Biennial I0CFR 50.59 and Commitment Revision Reports Summary of Justification:

Commitment Change Evaluation Form 10-07 revised the commitment due to the original stated procedures are no longer valid. Exelon procedure CC-AA-204 now ensures and governs the VETIP Program and Recontact Program for conformance to GL 90-03 and its supplement.

Letter Source: 5211-85-2160 "GL 83-28 ATWS Vendor Interface Manual Document Control NUTAC" Exelon Tracking No.: 1985T0060 / 1122355-07 Nature of Commitment: GPUN committed to the NRC to issue and maintain procedures that ensure vendor documents are properly reviewed for impact on existing documents and disseminated for action.

Summary of Justification:

Commitment Change Evaluation Form 10-07 revised the commitment due to the original stated procedures governing VETIP Program were voided. Exelon procedure CC-AA-204 now ensures and governs the VETIP Program and Recontact Program for conformance to GL 90-03 and its supplement (reference section 4 through 4.8.2).

Letter Source: C311-90-2073 "TMI Fire Brigade Training" Exelon Tracking No.: 1990T0050 / 1122355-57 Nature of Commitment: Each Fire Brigade member shall participate in at least two drills per year at TMI-1 to remain qualified.

Summary of Justification:

Commitment Change Evaluation Form 10-12 revised the commitment. The change provides individual brigade members credit for successfully performed fire drills regardless of location (either unit). Prior SER's for both TMI-1 and TMI-2 required 2 drills per year per unit per person.

In preparation for entry into TMI-2 Post Defueled Monitored Storage (PDMS), the requirement was changed to 2 per year in Unit 1 with no individual member requirement in TMI-2. The brigade leader was (and still is per section 4.11.5.1.3 and 4 of AP 1038) to perform 2 per year in Unit 1 and 1 per year in Unit 2. Brigade members must participate to support drill performance (not required for brigade member qualification). All fire drills at TMI regardless of location are performed per AP 1038 and implementing documents (see section 4.12). Drills are preplanned, have objectives, and are critiqued. Drills require the brigade to practice as a team, involve the use of equipment with very little simulation and involve the TMI-1 Control Room crew regardless of location on site. Drills within TMI-2 challenge operator knowledge of TMI-2 and are good practice (includes electrical distribution, radiation protection / releases, equipment operability, procedure and preplan knowledge and use, impact on the safe operation of TMI-1, etc. With the entry into PDMS, Unit 1 Operations and Maintenance became responsible for Unit 12

2010-2011 Biennial IOCFR 50.59 and Commitment Revision Reports 2 following phase-in which included training and transfer of key personnel to Unit 1 organizations. This provided considerable knowledge of the Unit 2 plant and PDMS configuration and requirements on the brigade. Therefore this change does not detract from the ability to practice as a team and be evaluated against drill objectives per the Fire Protection Program requirements.

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Repetto, John From: Julian, Emile Sent: Tuesday, April 17, 2012 4:57 PM To: Repetto, John Cc: Ngbea, Evangeline; Giitter, Rebecca

Subject:

RE: ML12108A114 in EHD Package ML12108A108 - Has Copyright Material The sensitivity would be non-public(also whatever the copyright one is) but yes please copy to the non-public folder.

From: Repetto, John Sent: Tuesday, April 17, 2012 4:33 PM To: Julian, Emile; Docket, Hearing; Greathead, Nancy; Giitter, Rebecca; Pierpoint, Christine; Ngbea, Evangeline Cc: Prater, Nancy

Subject:

RE: ML12108A114 in EHD Package ML12108A108 - Has Copyright Material

Emile, We can change the profile to Non-Public for that piece. What would the sensitivity be? Would it be Non-Sensitive or Sensitive-(indicate value)? When we are done with the package, do you want us to copy ML12108A114 over to Electronic Docket-Non-Public folder?

Thanks Rick From: Julian, Emile Sent: Tuesday, April 17, 2012 4:28 PM To: Repetto, John; Docket, Hearing; Greathead, Nancy; Glitter, Rebecca; Pierpoint, Christine; Ngbea, Evangeline Cc: Prater, Nancy

Subject:

RE: ML12108A114 in EHD Package ML12108A108 - Has Copyright Material

Rick, Thanks for bringing the attachment to my attention. There are two doctorial dissertations in the attachment that are copyrighted. Both are also on line with copyright notices there as well. I think we have no choice but to process ML12108A114 separately as a non-public document. The package can be made public with the other documents.

Emile From: Repetto, John Sent: Tuesday, April 17, 2012 2:24 PM To: Docket, Hearing; Julian, Emile; Greathead, Nancy; Gitter, Rebecca; Pierpoint, Christine Cc: Prater, Nancy

Subject:

ML12108A114 in EHD Package ML12108A108 - Has Copyright Material The subject package was received for DPC processing. The document, ML12108A114, has copyright mark on page 28. Is it okay for DPC to declare the documents and the container?

Thanks Rick Repetto (DPC) 1.

Date: 04/18/2012 DPC PROFILE ERROR LOG QC'er Name: Raymond Summers DPC PROFILERS Data Fields G. Newton C. Shipp A. Andrews A. Hale A. Walcott E. Shackelford D. Philpott Page Count Document Date Document Type Availability.

Title Author Name Author Affiliation Addressee Name Addressee Affiliation.-'

Docket Number License Number Case Ref. Number Document Report Number Keyword Release= Date Distribution List Code TOTALERRORS 0 0 0 1 0 1

Repetto, John From: Broadnax, Tawanna Sent: Wednesday, April 18, 2012 7:55 AM To: Repetto, John

Subject:

Security Timeline 8:00 8:45 9:30 10:15 11:00 11:45 12:30 1:15 2:00 2:45 3:30 4:00 1

Repetto, John From: Broadnax, Tawanna Sent: Wednesday, April 18, 2012 8:10 AM To: Repetto, John

Subject:

Package ML121080579 Container (Normal Processing Folder)

The subject package has one public document and one non-public document but the container is Public.

1