ML12072A429
| ML12072A429 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/23/2012 |
| From: | Tam P Plant Licensing Branch III |
| To: | Weber L Indiana Michigan Power Co |
| Tam P | |
| References | |
| TAC ME6882, TAC ME6883 | |
| Download: ML12072A429 (11) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 23, 2012 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT (CNP), UNITS 1 AND 2 - SAFETY EVALUATION FOR THE COMMITMENT RELATING TO THE ALLOY 600 AGING MANAGEMENT PROGRAM (TAC NOS. ME6882 AND ME6883)
Dear Mr. Weber:
On August 30, 2005, the CNP units' operating licenses were renewed, supported by the Nuclear Regulatory Commission (NRC) staffs Safety Evaluation Report dated May 2005. Appendix A of the Safety Evaluation Report affirmed the commitment made by the licensee:
The Alloy 600 Aging Management Program commitment will also be revised to indicate that an inspection plan will be submitted for staff review and approval three years prior to the period of extended operation to determine if the program demonstrates an ability to manage the effects of aging per 10 CFR 54.21 (a)(3).
By letter dated August 17,2011 (Accession No. ML11238A069), the licensee complied with the above commitment by submitting for NRC staff review and approval a document identified as EHI-5070-ALLOY 600, "Alloy 600 Material Management Program," Revision 3. By letter dated January 31, 2012 (Accession No. ML12048A072), the licensee submitted Revision 4 of the same document, and responded to the NRC staffs Request for Additional Information dated October 27 and November 3, 2011.
Based on its review set forth in the enclosed Safety Evaluation, the NRC staff finds that the licensee's EHI-50707-ALLOY600, "Alloy 600 Material Management Program," Revision 4, satisfies the NRC-approved Alloy 600 Aging Management Program (AMP) conveyed in the CNP license renewal application, as supplemented, the current requirements in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a, and guidance in Standard Review Plan for License Renewal, Section A.1.2.3, for nickel-based alloy components. Pursuant to 10 CFR 54.21 (a)(3), the NRC staff concludes that EHI-50707-ALLOY600, Revision 4, is acceptable because it is in concert with the NRC-approved Alloy 600 AMP, and it demonstrates that aging effects of the Alloy 600 material will be adequately managed so that the intended function
L. J. Weber
-2 of Alloy 600 components will be maintained consistent with the current licensing basis for the period of extended operation of CNP Units 1 and 2.
&rereIY'~~
~~, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosure:
Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT (CNP), UNITS 1 AND 2 ALLOY 600 AGING MANAGEMENT PROGRAM DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
By letter dated October 31, 2003 (Accession No. ML033070179), the licensee submitted its license renewal application for CNP, Units 1 and 2. In this application, among other things, was a commitment regarding Alloy 600 Aging Management Program. By letter dated August 11, 2004 (Accession No. ML042470410), and as result of a Request for Additional Information issued by the Nuclear Regulatory Commission (NRC) staff on July 2, 2004 (Accession No. ML041840194), the licensee conveyed a revised commitment, stating:
The Alloy 600 Aging Management Program commitment will also be revised to indicate that an inspection plan will be submitted for staff review and approval three years prior to the period of extended operation to determine if the program demonstrates an ability to manage the effects of aging per 10 CFR 54.21 (a)(3).
On August 30, 2005, the units' operating licenses were renewed (Accession No. ML052340447), supported by the NRC staff's Safety Evaluation Report dated May 2005 (Accession No. ML051510092). Appendix A of the Safety Evaluation Report affirmed the above cited commitment made by the licensee.
By letter dated August 17, 2011 (Accession No. ML11238A069), the licensee complied with the above cited revised commitment by submitting for NRC staff review and approval a document identified as EHI-5070-ALLOY 600, "Alloy 600 Material Management Program," Revision 3. By letter dated January 31, 2012 (Accession No. ML12048A072) the licensee submitted Revision 4 of the same document, and responded to the NRC staff's Request for Additional Information dated October 27 and November 3, 2011 (Accession Nos. ML113000267 and ML11307A485).
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR), Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," provides requirements for license renewal applications. Paragraph 10 CFR 54.21 (a)(3) requires that "... [f]or each structure and Enclosure
- 2 component identified in paragraph (a)(1) of this section, demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB [current licensing basis] for the period of extended operation..." The NRC staff reviewed the information included in the CNP License Renewal Application regarding the licensee's demonstration of the Alloy 600 Aging Management Program (AMP) to ensure that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the CLB throughout the period of extended operation.
The October 31,2003, license renewal application stated that the Alloy 600 AMP was a new AMP and there was no comparable AMP in NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Revision 0, at the time. In 2010, the NRC published the GALL report, Revision 2 (Accession No. ML101320104), which contains a new AMP, XI.M11B, "Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components (PWRs Only)."
3.0 NRC STAFF EVALUATION The plant-specific Alloy 600 AMP manages potential cracking in nickel-based Alloy 600 components and the corresponding weld materials resulting from primary water stress corrosion cracking (PWSCC) through assessment, inspection, mitigation, and repair or replacement of susceptible components. As documented in its Safety Evaluation Report of May 2005, the NRC staff approved the Alloy 600 AMP in the license renewal application to ensure that the aging effects of Alloy 600 nickel alloy components will be adequately managed so that the intended function of the Alloy 600 components will be maintained consistent with the current licensing basis for the period of extended operation. As part of its approval of the license renewal application, the NRC staff affirmed the licensee's commitment to submit an Alloy 600 inspection plan.
For the current submittal, the NRC staff reviews EHI-50707-ALLOY600, Alloy 600 Material Management Program, Revision 3, based on the same program description and ten program elements as discussed in the original Alloy 600 AMP in the license renewal application. The NRC staff verified that EHI-50707-ALLOY600 is based on the NRC-approved Alloy 600 AMP in the license renewal application, criteria defined in Standard Review Plan for License Renewal (SRP-LR), Section A.1.2.3, "Aging Management Elements," and the current requirements of 10 CFR 50.55a and 10 CFR 54.
3.1 Program Description Section 50.55a of 10 CFR requires the use of American Society of Mechanical Engineers (ASME) Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1." However, the NRC staff noted that Section 1.7 of Revision 3 of EHI-50707 ALLOY600 did not reference this code case. The NRC staff also noted that although Section 1.7 of EHI-50707-ALLOY600 identifies ASME Code Cases N722-1 and N-770-1, the program did not reference 10 CFR 50.55a(g)(6)(ii)(E) and 10 CFR 50.55a(g)(6)(ii)(F) which
- 3 impose conditions on both code cases. The NRC's position is that whenever these three ASME code cases are mentioned, corresponding 10 CFR 50.55a(g)(6)(ii)(O), (E), or (F) should also be mentioned because of the conditions imposed on these code cases.
The licensee responded by letter of January 31, 2012, stating that ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(O) address examination requirements for reactor vessel upper heads with nozzles having pressure-retaining partial-penetration welds and are not specifically applicable to Alloy 82/182 pipe butt welds. Therefore, it is not included in Section 1.7 of EHI 50707-ALLOY600. However, the licensee has revised Attachment 2 of EHI-50707-ALLOY600 to include a reference to 10 CFR 50.55a(g)(6)(ii)(O), 10 CFR 50.55a(g)(6)(ii)(E) and 10 CFR 50.55a(g)(6)(ii)(F).
The NRC staff finds that the revised EHI-50707-ALLOY600, Revision 4, referencing the appropriate 10 CFR 50.55a paragraphs with respect to the three ASME Code Cases; therefore, it satisfies the program description in the NRC-approved Alloy 600 AMP and SRP-LR, Section A.1.2.3.
3.1.1 Program Element No.1, Scope of the Program Sections 1 and 3 of EHI-50707-ALLOY600, Revision 3, describe the scope of the program for various nickel alloy component locations. Attachments 1,2,3 and 4 of EHI-50707-ALLOY600 provide specific details for each Alloy 600 component, locations, material specifications, inspection requirements (method, frequency, and acceptance criteria), susceptibility rankings, and mitigations and repairs.
Section 3.5.4 of EHI-50707-ALLOY600, Revision 3, states that u *** [d]etermine appropriate sample locations, inspection techniques, and acceptance standards in accordance with industry guidelines... " The NRC staff noted that in addition to industry guidelines, Section 3.5.4 needs to state that the sample locations, inspection techniques and acceptance standards are also determined in accordance with 1 0 CFR 50.55a and relevant NRC generic communications. The NRC staff requested that the licensee discuss the inspection samples for the components identified in Attachment 2 during each inspection.
The licensee responded in its January 31, 2012, letter that it has revised Section 3.5.4 of EHI 50707-ALLOY600, Revision 3, to state that "... sample locations, inspection techniques and acceptance standards are determined in accordance with industry guidelines, NRC regulations (10 CFR 50.55a), and relevant NRC generic communications." The licensee stated that of EHI-50707-ALLOY600, Revision 4, provides a list of inspection requirements for components in the Alloy 600 AMP. For the inspections that are required by 10 CFR 50.55a, the inspection samples are determined in accordance with the requirements provided in 10 CFR 50.55a (e.g., the ASME Code,Section XI and Code Cases). The required inspection samples for steam generator tube inspections are provided in the approved Technical Specifications (TS) for CNP.
The NRC staff finds that EHI-50707 -ALLOY600, Revision 4, satisfies program element No.1, "Scope of the Program," in the NRC-approved Alloy 600 AMP and criterion defined in SRP-LR Section A.1.2.3.1.
- 4 3.1.2 Program Element No.2, "Preventive Actions" to EHI-50707-AllOY600, Revision 4, provides information on mitigation and repairs for Alloy 600 components. Specifically, the licensee has replaced the reactor vessel closure head in CNP Units 1 and 2, applied mechanical stress improvement process on all eight hot leg and cold leg nozzles attached to the reactor vessel in CNP unit 1, and installed weld overlays on various pressurizer nozzles in CNP units 1 and 2. The NRC staff finds that the licensee has performed and implemented preventive actions to mitigate susceptible Alloy 600 components.
The NRC staff finds that EHI-50707-AllOY600, Revision 4, satisfies program element No.2, "Preventive Actions," in the NRC-approved Alloy 600 AMP and the guidance in SRP-lR Section A.1.2.3.2.
3.1.3 Program Element No.3, "Parameters Monitored or Inspected" of EHI-50707-AllOY600, Revision 3, specifies "Visual" as the inspection method for various steam generator components such as divider plate, tubesheet cladding, manway diaphragm, nozzle dam rings, and shell cladding. By its January 31, 2012, letter the licensee clarified that the visual inspection performed for steam generator components are not equivalent to any ASME Code examinations. These inspections consist of a visual scan of the listed component for any signs of degradation or cracking. The acceptance criteria for these inspections are met if no signs of degradation or cracking are observed. The NRC staff notes that revised EHI-50707-AllOY600, Revision 4, uses the appropriate volumetric, surface and visual nondestructive examination techniques for detection of degradation of the components identified in the scope as required by the ASME Code and industry guidance.
The NRC staff finds that EHI-50707-AllOY600, Revision 4, satisfies program element No.3, "Parameters Monitored or Inspected," in the NRC-approved Alloy 600 AMP and guidance in SRP-lR Section A.1.2.3.3.
3.1.4 Program Element No.4, "Detection of Aging Effects" As part of detection of aging effects, the NRC staff asked the licensee to discuss the capability of the reactor coolant system (RCS) leakage detection systems in CNP, Units 1 and 2, with respect to Regulatory Guide 1.45, Revision 1, "Guidance on Monitoring and Responding to Reactor Coolant System leakage." By its January 31,2012, letter the licensee stated that the construction permits for CNP were issued and the majority of construction was completed prior to issuance of 10 CFR 50, Appendix A, General Design Criteria, in 1971, by the Atomic Energy Commission (AEC). CNP was designed and constructed to comply with the AEC General Design Criteria (GDC) as proposed on July 10, 1967. The application of the AEC proposed General Design Criteria to CNP is contained in the CNP Updated Final Safety Analysis Report (UFSAR) as the Plant Specific Design Criteria (PSDC). Appendix A of 10 CFR 50 GDC differs both in numbering and content from the PSDC for CNP.
The licensee stated that PSDC 16, "Monitoring Reactor Coolant leakage," describes the means that are provided to detect Significant uncontrolled leakage from the reactor coolant pressure
- 5 boundary. This requirement meets the intent of GDC 30, which requires that means be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.
The licensee stated that CNP is not committed to Regulatory Guide (RG) 1.45. However, the requirements of RG 1.45 were followed to the extent practical in the design of the leakage monitoring systems. A detailed discussion of RCS leakage monitoring is provided in CNP UFSAR Section 4.2.7, "Leakage."
The Unit 1 containment atmosphere particulate radioactivity monitor has a licensing basis leak detection capability of 0.8 gallons per minute (gpm) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Unit 1 containment atmosphere gaseous radioactivity monitor has a licensing basis leak detection capability of 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The Unit 2 containment atmosphere particulate and gaseous radioactivity monitors have a licensing basis leak detection capability of 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The NRC staff previously reviewed and approved the above CNP's leakage detection capability in License Amendment Nos. 317 (for Unit 1) and 300 (for Unit 2), dated November 1, 2011 (Accession No. ML11249A090). These amendments revised TS 3.4.15, "RCS Leakage Detection Instrumentation," and reviewed the associated TS Bases pages consistent with Revision 3 of Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-513, "Revise PWR Operability Requirements and Actions for RCS Leakage Instrumentation."
The NRC staff finds that EH 1-50707 -ALLOY600, Revision 4, satisfies program element No.4, "Detection of Aging Effects," in the NRC-approved Alloy 600 AMP and guidance in SRP-LR Section A.1.2.3.4.
3.1.5 Program Element No.5, "Monitoring and Trending" For inservice inspection (lSI), EHI-50707-ALLOY600, Revision 4, follows 10 CFR 50.55a, the ASME Code Section XI, and NRC staff-accepted industry guidance for monitoring and trending of nickel alloy components. ASME Section XI requires (a) recording of examination and test results that provide a basis for evaluation and facilitate comparison with the results of subsequent examinations, (b) retention of all inspection, examination, test, and repair/replacement activity records and flaw evaluation calculations for the service lifetime of the component or system, and (c) additional examinations (Le., sample expansion), when flaws or relevant conditions are found that exceed the applicable acceptance criteria, to assist in determination of an extent of condition and causal analysis.
The NRC staff finds that EHI-50707-ALLOY600, Revision 4, satisfies program element No.5, "Monitoring and Trending," in the NRC-approved Alloy 600 AMP and guidance in SRP-LR Section A.1.2.3.5.
3.1.6 Program Element No.6, "Acceptance Criteria" Section 4.7.1 of EHI-50707 -ALLOY600, Revision 3, states that"... [i]nspection results that do not meet the acceptance criteria are documented via PMP-7030-CAP-001, Action Initiation... " The NRC staff noted that the proposed inspection plan needs to include a section on acceptance
-6 criteria for the inspection results and the acceptance criterion for individual components should be defined. By its January 31, 2012, letter the licensee provided acceptance criteria in Section 4.7 of EHI-50707-ALLOY600, Revision 4. The licensee also revised Attachment 2 of EHI-50707 -ALLOY600 to list the specific acceptance criteria for each inspection. The NRC staff finds that the revised Attachment 2 in EHI-50707-ALLOY600, Revision 4, includes appropriate acceptance criteria for each Alloy 600 components and, therefore, is acceptable.
The NRC staff finds that EHI-50707-ALLOY600, Revision 4, satisfies program element No.6, "Acceptance Criteria," in the NRC-approved Alloy 600 AMP and the guidance in SRP-LR Section A.1.2.3.6.
3.1.7 Program Element No.7, "Corrective Actions" Section 4.7.5 of EHI-50707-ALLOY600, Revision 3, states that "... [b]ased on the initial inspection results, the need for additional inspections are determined. This information is used to develop future inspection scope and associated inspection intervals. Subsequent inspections may include inspections of the additional locations." The NRC staff asked the licensee to discuss the scope and intervals of the additional inspection with respect to the ASME Code,Section XI, if the initial inspection discovers degradation in Alloy 600 components.
The licensee responded in its January 31,2012, letter that Section 4.7.5 of EHI-50707 ALLOY600, Revision 3, was deleted and Attachment 2 was revised to provide a list of inspection requirements for Alloy 600 components. The licensee clarified that, as required by 10 CFR 50.55a, several of these inspections are conducted in accordance with ASME Code,Section XI, or applicable ASME Code Cases. The licensee will evaluate the results of these inspections in accordance with the applicable ASME Code and 10 CFR 50.55a to determine if any additional inspections are required. If additional inspections are required, the licensee will implement those inspections in accordance with the applicable ASME Code and 10 CFR 50.55a requirements. The NRC staff finds that the revised Attachment 2 in EHI-50707-ALLOY600, Revision 4, is acceptable because the inspection guidance in Attachment 2 follows the latest 10 CFR 50.55a requirements.
The NRC staff finds that EHI-50707-ALLOY600, Revision 4, satisfies program element No.7, "Corrective Actions," in the NRC-approved Alloy 600 AMP and guidance in SRP-LR Section A.1.2.3.7.
3.1.8 Program Element No.8, "Confirmation Process" Section 4.7.2 of EHI-50707-ALLOY600, Revision 3, states that 'WHEN the test acceptance criteria are not met, THEN an Engineering Evaluation is performed in accordance with 10 CFR Part 50, Appendix B, and documented in accordance with PMP-7030-CAP-002, Condition Evaluation, Action, and Closure, in order to verify that the intended functions of the in-scope components can be maintained consistent with the current licensing basis... "
The NRC questioned why 1 0 CFR Part 50, Appendix B, is used as the basis for the engineering evaluation, in lieu of 10 CFR 50.55a which requires the use of the evaluation in the ASME Code,Section XI. In addition, the NRC staff asked the licensee to discuss the procedures that will be used to implement the requirements of 10 CFR Part 50, Appendix B, and to discuss whether the engineering evaluation would include corrective actions.
-7 The licensee responded in its January 31,2012, letter that it has revised Section 4.8.2
[previously 4.7.2], Corrective Measures. of EHI-S0707-ALLOY600, Revision 4, which now states: "WHEN the acceptance criteria are not met, THEN an evaluation is performed in accordance with 10 CFR SO.SSa, ASME Boiler and Pressure Vessel Code (Section XI or applicable Code Case), and industry guidelines."
The licensee explained that inspection results that do not meet the acceptance criteria are documented and addressed in accordance with the Corrective Action Program, which is compliant with 10 CFR SO, Appendix B. Criterion XVI, "Corrective Action." Corrective action procedures are implemented in accordance with the requirements of 10 CFR SO, Appendix B.
The licensee stated that when acceptance criteria are not met, an evaluation will be performed in accordance with 10 CFR SO.SSa, ASME Code (Section XI or applicable Code Case), and industry guidelines. The evaluation will determine if the component is acceptable for continued service and the corrective actions required (e.g., further evaluation, additional examinations, increased inspection frequency, and repair/replacement). The NRC staff finds that the licensee has revised and clarified the corrective actions in the proposed inspection plan with respect to the requirements in 10 CFR SO, Part Band 10 CFR SO.SSa.
The NRC staff finds that EHI-S0707-ALLOY600, Revision 4, satisfies program element No.8, "Confirmation Process," in the NRC-approved Alloy 600 AMP and the guidance in SRP-LR Section A.1.2.3.8.
3.1.9 Program Element No.9, "Administrative Controls" The licensee stated that it implements the CNP quality assurance procedures, review and approval processes, and administrative controls in accordance with the requirements of Appendix B to 10 CFR Part SO. The licensee performs administrative control for both safety related and nonsafety-related structure and components according to the existing Document Control Program in accordance with the Quality Assurance Program Description. The licensee stated that the CNP administrative controls are consistent with the GALL Report.
The NRC staff finds that EHI-S0707-ALLOY600, Revision 4, satisfies program element No.9, "Administrative Controls," in the NRC-approved Alloy 600 AMP and guidance in SRP-LR Section A.1.2.3.9.
3.1.10 Program Element No.1 0, "Operating Experience" The licensee discussed the indications detected in one of the control rod drive mechanism nozzles in the reactor vessel head at Unit 2 in 1994 refueling outage. The licensee replaced the Unit 2 reactor vessel head in 2007. The licensee discussed indications detected at the pressurizer safety nozzle at Unit 1 in 200S. The licensee repaired the nozzle using full structural weld overlay. The licensee lists its response to the NRC generic communications regarding cracking in Alloy 600 material.
The NRC staff finds that EHI-S0707-ALLOY600, Revision 4, satisfies program element No. 10, "Operating Experience," in the NRC-approved Alloy 600 AMP and guidance in SRP-LR Section A.1.2.3.10.
- 8
4.0 CONCLUSION
On the basis of its review set forth above, the NRC staff finds that the licensee's EHI-50707 ALLOY600, "Alloy 600 Material Management Program," Revision 4, satisfies the NRC-approved Alloy 600 AMP conveyed in the CNP license renewal application, as supplemented, the current requirements in 10 CFR 50.55a, and guidance in SRP-LR, Section A.1.2.3, for nickel-based alloy components. Pursuant to 10 CFR 54.21 (a)(3), the NRC staff concludes that EHI-50707 ALLOY600, Revision 4, is acceptable because it is in concert with the NRC-approved Alloy 600 AMP, and it demonstrates that aging effects of the Alloy 600 material will be adequately managed so that the intended functions of Alloy 600 components will be maintained consistent with the current licensing basis for the period of extended operation of CNP Units 1 and 2.
Principal Contributor: John Tsao Date: March 23, 2012
- 2 L. J. Weber of Alloy 600 components will be maintained consistent with the current licensing basis for the period of extended operation of CNP Units 1 and 2.
Sincerely, lRAI Peter S. Tam, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosure:
Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 Resource RidsNrrPMDCCook Resource RidsNrrLABTully Resource RidsOgcRp Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDssStsb Resource RidsNrrDorlDprResource RidsRgn3MailCenter Resource RidsNrrDssSnpb Resource RidsNrrDeEpnb Resource J. Tsao, NRR ADAMS Accession No: ML12072A429
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LPL3-1/PM NAME PTam BTuily
- TLupold*
~ SWiliiams PTam DATE 3/20/12 3/20/12 I 3/1112 I 3/21/12
- 3/23/12
- Safety evaluation transmitted by memo of 311/12 (Accession No. ML12065A115).
OFFICIAL RECORD COpy