ML15062A013

Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments 213 and 201 to Revise TS 3.5.3, ECCS - Shutdown (TAC Nos. MF0727 and MF0728)

ML15062A013
Person / Time
Site:	Prairie Island
Issue date:	05/20/2015
From:	Beltz T A Plant Licensing Branch III
To:	Davison K K Northern States Power Co
Beltz T A
References
TAC MF0727, TAC MF0728
Download: ML15062A013 (18)

Text

Mr. Kevin K. Davison Site Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 20, 2015 Prairie Island Nuclear Generating Plant Northern States Power Company -Minnesota 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1AND2 -ISSUANCE OF LICENSE AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATION 3.5.3, "EGGS-SHUTDOWN" (TAC NOS. MF0727 AND MF0728)

Dear Mr. Davison:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 213 and 201 to Renewed Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated February 20, 2013, as supplemented by letters dated June 25, 2013, September 15, 2014, and February 26, 2015. The amendments revise TS 3.5.3, "ECCS -Shutdown." The amendments remove Note 1 and change the Mode Applicability to eliminate the potential for non-conservative plant operation.

K. Davison A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket Nos. 50-282 and 50-306

Enclosures:

1. Amendment No. 213 to DPR-42 2. Amendment No. 201 to DPR-60 3. Safety Evaluation cc: Distribution via Listserv Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY -MINNESOTA DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 213 Renewed License No. DPR-42 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northern States Power Company -Minnesota (NSPM, the licensee), dated February 20, 2013, as supplemented by letters dated June 25, 2013, September 15, 2014, and February 26, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act}, and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (1 O CFR) Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-42 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications. Enclosure 1 3. This license amendment is effective as its date of issuance and shall be implemented within 90 days from the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. DPR-42 and Technical Specifications Date of Issuance: May 20, 2015 FOR THE NUCLEAR REGULA TORY COMMISSION David . Pel on, hie Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 213 RENEWED FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 Replace the following page of Renewed Facility Operating License No. DPR-42 with the attached revised page. The change area is identified by a marginal line. REMOVE INSERT Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change. REMOVE INSERT 3.5.3-1 3.5.3-1 (3) Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 213, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications. (3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-42 Amendment No. 213 ECCS -Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS -Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. -------------------------------------NOTE------------------------------------------An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room. APPLICABILITY: MODE 4. ACTIONS -------------------------------------------------NOTE------------------------------------------------------LCO 3.0.4.b is not applicable to ECCS safety injection (SI) subsystem. CONDITION A. Required ECCS residual heat removal (RHR) subsystem inoperable. B. Required ECCS safety injection (SI) subsystem inoperable. Prairie Island Units 1and2 A.1 B.1 REQUIRED ACTION COMPLETION TIME Initiate action to restore Immediately required ECCS RHR subsystem to OPERABLE status. Restore required ECCS SI lhour subsystem to OPERABLE status. 3.5.3-1 Unit 1 -Amendment No. f69. 1-83-213 Unit 2 -Amendment No. 1-51-¥B 201 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY -MINNESOTA DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 201 Renewed License No. DPR-60 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northern States Power Company -Minnesota (NSPM, the licensee), dated February 20, 2013, as supplemented by letters dated June 25, 2013, September 15, 2014, and February 26, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-60 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 201, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications. Enclosure 2 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. DPR-60 and Technical Specifications Date of Issuance: May 20, 2015 FOR THE NUCLEAR RE7TORY COMMISSION £) 1{ . . Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 201 RENEWED FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 Replace the following page of Renewed Facility Operating License No. DPR-60 with the attached revised page. The change area is identified by a marginal line. REMOVE INSERT Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change. REMOVE INSERT 3.5.3-1 3.5.3-1 (3) Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 1 O CFR Chapter I: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 201, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications. (3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 201 ECCS -Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS -Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. -------------------------------------NOTE------------------------------------------An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room. APPUCABil.1TY: MODE 4. ACTIONS -------------------------------------------------NOTE------------------------------------------------------LCO 3.0.4.b is not applicable to ECCS safety injection (SI) subsystem. ---------------------------------------------------------------------------------------------------------------CONDITION A. Required ECCS residual heat removal (RHR) subsystem inoperable. B. Required ECCS safety injection (SI) subsystem inoperable. Prairie Island Units 1and2 . A.I B.l REQUIRED ACTION COMPLETION TIME Initiate action to restore Immediately required ECCS RHR subsystem to OPERABLE status. Restore required ECCS SI 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> subsystem to OPERABLE status. 3.5.3-1 Unit 1 -Amendment No. 167 83-213 Unit 2 -Amendment No. f5.'.t-¥B 201 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 213 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 201 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY -MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated February 20, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13053A 199), as supplemented by letters dated June 25, 2013, September 15, 2014, and February 26, 2015 (ADAMS Accession Nos. ML 13178A024, ML 14258A089, and ML 15057 A480, respectively), Northern States Power Company -Minnesota (NSPM, the licensee), doing business as Xcel Energy, Inc., requested changes to the technical specifications (TSs) for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. The licensee proposed a revision to TS 3.5.3, "ECCS [Emergency Core Cooling] -SHUTDOWN," to remove Limiting Condition for Operation (LCO), Note 1, which was identified by the licensee as being incorrect and may not adequately maintain system operability. The licensee had been administratively controlling plant operations in TS 3.5.3 under the provision of U.S. Nuclear Regulatory Commission (NRC) Administrative Letter (AL) 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," and submitted its license amendment request in accordance with the guidance provided in AL 98-10. In its letter dated February 26, 2015, the licensee proposed to change the TS 3.5.3 Applicability to "MODE 4," removing the currently defined Applicability which states "MODE 4 when both RCS [reactor coolant system] cold leg temperatures are > [greater than] SI [safety injection] pump disable temperature specified in PTLR [Pressure Temperature Limits Report]." The supplements dated June 25, 2013, and September 15, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant Enclosure 3 hazards consideration determination as published in the Federal Register on August 20, 2013 (78 FR 51229). A revised no significant hazards consideration determination was published in the Federal Register on March 17, 2015 (80 FR 13910), to consider the additional aspects of the proposed Applicability change provided in the February 26, 2015, supplemental letter.

2.0 REGULATORY EVALUATION

The Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, was licensed to operating prior to issuance of the regulations in Appendix A, "General Design Criteria [GDC] for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50. PINGP was licensed on the basis of similar plant-specific principal design criteria1. PINGP was licensed on the basis of similar plant-specific principal design criteria. Section 1.5, "General Design Criteria, of the PINGP Updated Safety Analysis Report indicates that PINGP was designed and constructed to comply with the licensee's understanding of the intent of Atomic Energy Commission's GDCs for nuclear power plant construction permits as proposed on July 10, 1967. The Appendix A requirements similar to the Pl NGP principal design criteria include the following:

• GDC 34 requires a residual heat removal (RHR) system designed to maintain specified acceptable fuel design limits and ensure the design conditions of the reactor coolant boundary are not exceeded if a single failure occurs simultaneous with failure of specified electrical power systems.

• GDCs 35, 36, and 37 require an ECCS design that meets certain performance, inspection, and testing requirements. The regulation in 10 CFR 50.36(c)(2) defines an LCO as the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a LCO of a nuclear reactor is not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The regulation in 10 CFR 50.46 provides the ECCS acceptance criteria for light-water nuclear power reactors. The regulations specify requirements of ECCS to ensure that the calculated cooling performance criteria of 10 CFR 50.46, paragraph (b), are met following postulated LOCAs.

3.0 TECHNICAL EVALUATION

The primary function of the ECCS is to remove heat from the reactor core during accident conditions. During the initial part of an event requiring ECCS actuation at PINGP, a flow path is required to provide refueling water storage tank (RWST) water to the RCS. For long-term cooling, RCS make-up water may be supplied from the containment sump. Following a loss-of-1 These apply to facilities with construction permits issued before May 21, 1972, and that were not licensed under 10 CFR Part 50, Appendix A. Prairie Island Unit 1 was licensed to operate on*August 9, 1973, and Unit 2 was licensed to operate on October 29, 1974. The operating licenses for both units were renewed on June 27, 2011.

coolant accident (LOCA), the RHR system performs two functions as the low-head portion of the ECCS:

• When RCS pressure is low enough, it provides borated water from the RWST into the RCS, and

• It provides a long-term recirculation capability from the containment sump for core cooling. During Modes 1, 2, and 3, both ECCS trains are required to be operable and the RHR system is capable of being aligned to perform the ECCS low-head injection function. During Mode 4, the TS 3.5.3 LCO requires that one ECCS train be operable. The RHR pumps may also be used for core decay heat removal during normal reactor heat removal when in Modes 4, 5, and 6. TS 3.5.3 requires an ECCS train to be operable during Mode 4, where the RCS average coolant temperature, Tave. is described by a range of 200 degrees Fahrenheit (°F) < Tave< 350°F. The RHR system is part of the ECCS. If an RHR train has been operating to remove heat from the RCS, a portion of the RHR system will be subject to RCS hot leg temperature, T hot* Transferring an RHR train to an ECCS configuration when T hot is greater than or equal to the local saturation temperature in the RHR piping, Tsat. may cause a loss of ECCS capability due to flashing in the RHR piping. The current PINGP TS 3.5.3 reads as follows: 3.5.3 ECCS -Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. --------------------------------------------NOTES ----------------------------------1. An RHR train may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation. 2. An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room. APPLICABILITY: MODE 4 when both RCS cold leg temperatures are> SI pump disable temperature specified in PTLR. Note 1 allows both RHR trains to be in operation in the shutdown cooling mode during Mode 4 when both RCS cold leg temperatures are greater than the SI pump disable temperature specified in PTLR. As stated in the February 20, 2013, submittal, the SI pump disable temperature currently corresponds to a value of 218°F. The licensee proposed to remove Note 1 to avoid any potential loss of ECCS capability due to flashing in the RHR piping. The licensee also proposed an enhancement to TS 3.5.3 by expanding the mode applicability to cover all of Mode 4, in lieu of the existing TS 3.5.3 mode applicability which specifies a lower range defined when RCS cold leg temperature is above the SI disable temperature specified in the PTLR (i.e., 218°F). Based on the above, the licensee proposed to revise TS 3.5.3 to read as follows: 3.5.3 ECCS -Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE. -------------------------------------------l\IOTE -------------------------------An SI train may be considered OPERABLE when the pump is capable of being manually started from the control room. APPLICABILITY: MODE 4 The l\IRC staff finds that removing l\lote 1 from TS 3.5.3 is acceptable, as it lessens the potential for steam flashing and loss of pump suction and reduces the potential for a condition that could result in a loss of operability. The l\IRC staff finds that removing the qualifier, "when both RCS cold leg temperatures are greater than SI pump disable temperature specified in PTLR," provides a significant improvement to the TSs, as it provides additional assurance of continued ECCS availability that is particularly important when there exists the possibility of flashing during a loss-of-coolant or loss of RCS pressure, and is acceptable. The aforementioned changes address actual conditions in the hot leg that are of concern with respect to maintaining RHR system operability and furthering the capability to meet the ECCS acceptance criteria as specified in 10 CFR 50.46. Summary Based on the above, the l\IRC staff concludes that the revisions to TS 3.5.3 improve the capability for the RHR system trains to remain available through Mode 4 without the potential for steam flashing and loss of safety function. Therefore, the staff finds that the revised TS is consistent with the requirements of GDCs-34, 35, 36, and 37, as well as 1 O CFR 50.46. 4.0 STATE COl\ISUL TATIOl\I In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendment on April 15, 2015. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the use of facility components located within the restricted area as defined in 10 CFR Part 20 or changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding as published in the Federal Register on August 20, 2013 (78 FR 51229), and March 17, 2015 (80 FR 13910). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.

6.0 CONCLUSION

The NRC staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: Warren Lyon, NRR Date: May 20, 2015 K. Davison A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket Nos. 50-282 and 50-306

Enclosures:

Sincerely, IRA/ Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1. Amendment No. 213 to DPR-42 2. Amendment No. 201 to DPR-60 3. Safety Evaluation cc: Distribution via Listserv DISTRIBUTION: PUBLIC LPL3-1 r/f RidsNrrDorllpl3-1 Resource RidsRgn3MailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsNrrPMPrairielsland Resource RidsNrrLAMHenderson Resource RidsOgcRp Resource RidsNrrDorlDpr Resource ADAMS Accession No.: ML 15062A013 RidsNrrDssSrxb Resource RidsNrrDssStsb Resource Wlyon, NRR MHamm, NRR OFFICE NRR/DORL/LPL3-1 /PM N RR/DORL/LPL3-1 /LA NRR/DSS/SRXB/BC NRR/DSS/STSB/BC NAME TBeltz MHenderson CJackson RElliott DATE 04/03/2015 03/27/2015 03/03/2015 04/01/2015 OFFICE OGC (NLO, if issued after r 60-dav notice period) NRR/DORL/LPL3-1/BC NRR/DORL/LPL3-1 /PM .. // NAME MYounQ DPelton TBeltz """ DATE 04/14/2015 05/01/2015 05/20/2015 .. .,i;*< ... ;* OFFICIAL RECORD COPY