ML101580422

Relief 09-MN-005 for Alternative Leakage Testing for Various American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1 Piping and Components During the Third 10-Year Inservice

ML101580422
Person / Time
Site:	McGuire, Mcguire
Issue date:	06/14/2010
From:	Kulesa G Plant Licensing Branch II
To:	Repko R Duke Energy Carolinas
Thompson Jon, NRR/DORL/LPL 2-1, 415-1119
References
09-MN-005, TAC ME1732, TAC ME1733
Download: ML101580422 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 14, 2010 Mr. Regis T. Repko Vice President McGuire Nuclear Station Duke Energy Carolinas, LLC 12700 Hagers Ferry Road Huntersville, NC 28078 SUB..IECT:

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - RELIEF 09-MN-005 FOR ALTERNATIVE LEAKAGE TESTING FOR VARIOUS AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME), BOILER AND PRESSURE VESSEL CODE (CODE), CLASS 1 PIPING AND COMPONENTS DURING THE THIRD 10-YEAR INSERVICE INSPECTION (lSI) INTERVAL (TAC NOS. ME1732 AND ME1733)

Dear Mr. Repko:

By letter dated June 16, 2009, Duke Energy Carolinas LLC (the licensee), submitted relief request (RR) 09-MN-005 for McGuire Nuclear Station, Units 1 and 2, (McGuire 1 and 2) related to the third 10-year lSI interval pertaining to utilization of alternative leakage testing for various ASME Code,Section XI, Class 1 piping and components. The third 10-year lSI interval ends on December 1, 2011, for McGuire 1 and July 14, 2014, for McGuire 2. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(ii), the licensee requested to use an alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty. The licensee requested approval of a proposed alternative to the ASME Code requirements for system leakage testing for various ASME Code Class 1 piping and components connected to the reactor coolant system. In lieu of the ASME Code requirement to pressurize all Class 1 pressure retaining components within the system boundary to the code-required pressure, the licensee has proposed an alternative to conduct the system pressure test for certain Class 1 segments at a reduced pressure.

However, the visual examination during the pressure would include all components within the system boundary.

The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's submittal and concludes, as set forth in the enclosed Safety Evaluation, that the proposed alternative provides reasonable assurance of the structural integrity of the subject components and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR Section 50.55a(a)(3)(ii), the !\\IRC staff authorizes the use of the proposed alternative to conduct the system pressure test for certain Class 1 segments at a reduced pressure. However, the visual examination during the pressure would include all components within the system boundary.

R. Repko

- 2 All other requirements of ASME Code,Section XI for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Sincerely, loria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-369 and 50-370

Enclosure:

Relief cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THIRD 10-YEAR INTERVAL INSERVICE INSPECTION RELIEF NO. 09-MN-005 DUKE ENERGY CAROLINAS, LLC MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated June 16, 2009 (Agencywide Documents Access and Management System Accession No. ML091770549), Duke Energy Carolinas, LLC, (the licensee), submitted Relief Request (RR) No. 09-MN-005, related to the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, requirements for the third 10 year interval inservice inspection (lSI) program for the McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2). In RR 09-MN-005, the licensee proposed alternate pressure testing criteria during performance of the system pressure test for the Class 1 piping segments in the reactor coolant system for the pressurizer auxiliary spray, low pressure and high pressure safety injection, residual heat removal, and reactor coolant vents and drain piping. In lieu of the ASME Code requirement to pressurize all Class 1 pressure retaining components within the system boundary to the code-required pressure, the licensee has proposed alternatives to conduct the system pressure test for those Class 1 segments at a reduced pressure. However, the visual examination during the pressure test would include all components within the system boundary.

The licensee's request for relief is based on the hardship of performing off-normal activities in order to pressurize the portion of piping between the inboard and outboard isolation valves to the ASME Code Class 1 system leakage test pressure corresponding to 100% rated reactor power. Therefore, the Nuclear Regulatory Commission (NRC) staff has evaluated the lSI program alternatives proposed in RR 09-MN-005, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.55a(a)(3)(ii), for the third 1O-year lSI interval of McGuire 1 and 2.

2.0 REGULATORY EVALUATION

The regulation at 10 CFR 50.55a(g) requires that lSI of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). According to 10 CFR 50.55a(a)(3), alternatives to the requirements of Section 50.55a(g) may be used, when authorized by the NRC, if an applicant demonstrates that the Enclosure

- 2 proposed alternatives would provide an acceptable level of quality and safety, or if the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The regulations require that lSI of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The lSI Code of Record for the third 10-year lSI interval for McGuire 1 and 2 is the 1998 Edition of the ASME Code,Section XI with 2000 Addenda.

3.0 TECHNICAL EVALUATION

3.1 The Licensee's Relief Request In its letter dated June 16, 2009, the licensee requested relief from performing the system leakage test at a pressure corresponding to 100% rated reactor power. The licensee proposed alternative pressure testing criteria in lieu of the system leakage test required under IWB 5221 (a) for the above systems and components identified in Section 3.1.1.

3.1.1 Systems and Components for Which Relief is Requested Systems for which relief is requested Pressurizer Auxiliary Spray Reactor Head Vent Reactor Coolant Drains Low Pressure Safety Injection High Pressure Safety Injection Residual Heat Removal (RHR, or NO)

Components for which relief is requested Relief is requested for portions of ASME Code Class 1 piping and components connected to the reactor coolant system (RCS, or NC) that are normally isolated from direct RCS pressure (2235 pounds per square inch gauge (psig)) during their normal operation. They are isolated from the reactor coolant loop by their location: either upstream of a check valve, between two check valves, or between two closed valves that must remain closed during the unit's operation (or startup) in Modes 3, 2, or 1. The specific portions of piping for which relief is requested are described below.

- 3 Note: Valve/component numbers (even where unit numbers are listed) correspond for both McGuire 1 and 2.

Portion 1:

Portion 2:

Portion 3:

Portion 4:

Portion 5:

2-inch chemical volume and control system (NV) Class 1 piping and components upstream of auxiliary spray inboard check valve NV-22 up to and including outboard RCS isolation valves NV-21A (globe valve) and NV-841 (check valve).

14-inch and %-inch Class 1 piping and components on the NO suction line between the RCS double-isolation gate valves 1ND-01 Band 1ND 02AC (up to and including their gates).

On each of the 4 NC Loops, 1.5-inch safety injection system (NI) Class 1 piping and components between double-isolation check valves (and including the second isolation check valves) for NC cold leg boron injection. Double-isolation check valve pairs are:

NI-15 and NI-354 for Loop 1, NI-17 and NI-347 for Loop 2, NI-19 and NI-348 for Loop 3, NI-21 and NI-349 for Loop 4 On each of the 4 NC loops, 1O-inch, 6-inch, 2-inch, and %-inch NI Class 1 cold leg injection piping and components upstream of the 10-inch NC isolation check valves, and going back to and including the following:

a)

Cold leg accumulator isolation "block valve" (gate valve), the %

inch piping flow element and vent valves b)

NI pump and NO pump discharge isolation check valves (and associated %-inch piping):

NI-171 and NI-175 for Loop 1, NI-169 and NI-176 for Loop 2, 1\\11-167 and NI-180 for Loop 3, NI-165 and NI-181 for Loop 4 c)

NI-60, NI-71, NI-82, and NI-94 d)

NI-362, NI-364, NI-366, (McGuire 1) and inline flow restrictor (McGuire 2) 8-inch, 6-inch, 2-inch, 1-inch, and %-inch Class 1 piping and components in the NI upstream of the hot leg injection isolation check valves 1NI 157, 1NI-134, 1NI-126, and 1NI-160 (for Loops 1, 2, 3, and 4 respectively) and back to and including the following:

- 4 a)

NI pump 1B discharge isolation check valve 1NI-156 and associated %-inch line with flow restrictor (for Loop 1).

b)

NI pump 1A discharge isolation check valve 1NI-128 and NO pump(s) discharge isolation check valve 1NI-129 and associated

%-inch line with flow restrictor (for Loop 2).

c)

NI pump 1A discharge isolation check valve 1NI-124 and NO pump(s) discharge isolation check valve 1N1-125 and associated

%-inch line with flow restrictor (for Loop 3).

d)

NI pump 1B discharge isolation check valve 1NI-159 and associated %-inch line with flow restrictor (for Loop 4).

e)

Globe valves NI-423, NI-370, NI-368, (McGuire 1) and flow restrictors (McGuire 2).

Portion 6:

2-inch and 1-inch NC Class 1 piping and components between (and including) double-isolation globe valves isolating NC loop from the liquid waste recycle system piping routed to reactor coolant drain tank pump.

(One segment is on each of four Loops). Segment boundaries are:

NC-4, NC-5 and test drain NC-224 for Loop 1, NC-94, NC-95 and test drain NC-113 for Loop 2, NC-13, NC-106 and test drain NC-115 for Loop 3, NC-19, NC-20, 1NC-253 and test drain NC-111 for Loop 4.

NI-226 thru NI-229, NI-362 thru NI-371, NI-418 and NI-423 (McGuire 1)

Portion 7:

The Class 1 piping and components between (and including) valves on double-vent and/or double-drain valve assemblies installed on Class 1 piping headers. Also, the 1-inch NC piping between (and including) the following RCS double-isolation valve pairs on the reactor vessel head vent line:

NC-272 and NC-273, NC-274 and NC-275, NC-22 and NC-238.

3.1.2 Applicable ASME Code Requirements IWB-2500, Table IWB-2500-1, Examination Category B-P, Item Number B15.10, requires that all Class 1 pressure-retaining components be VT-2 visually examined each refueling outage.

The required system pressure test can be either a system hydrostatic test or a system leakage test. The system leakage test is performed at a pressure not less than the pressure corresponding to 100% rated reactor power. Per IWB-5222(a), the pressure-retaining boundary during the system leakage test shall correspond to the reactor coolant boundary with all valves

- 5 in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity. Per IWB 5222(b), the pressure-retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall be extended to all Class 1 pressure-retaining components within the system boundary.

3.1.3 Licensee's Basis for Requesting Relief The following discussion provides the basis for the requested relief and approval of the proposed alternative testing in accordance with the provisions of 10 CFR 50.55a(a)(3)(ii) due to the hardship that would be imposed by complying with the Code requirement.

Applying RCS operating pressure (2235 psig) to Portions 1, 2, 3, 6, and 7 of the Class 1 piping of this document would result in a hardship by exposing station personnel to:

personal safety hazards ranging from immediate physical exposure to temporary connctions whose medium is pressurized to 2235 psig (and in some cases at 557 F) to their being "stationed" at opened manual valves in lower containment at or near vent/drain valves serving as RCS single isolation pressure and temperature barriers in order to maintain the RCS boundary redundant valve protection requirement of 10 CFR 50.55a(c)(ii) during the test.

additional radiation exposure from activities in Lower Containment such as transporting, connecting, performing testing activities with, and removing hydro pump or temporary jumper materials; scaffold erection and tear down where needed; insulation removal and replacement where needed; valve internals removed and replaced where needed; and valve gags installed and removed where needed. Unknown delays in any of these activities could occur in lower containment, which would increase the additional radiation exposure.

Introducing NC system operating pressure (2235 psig) to the Portion 1 piping upstream of the pressurizer isolation check valve NV-22 during Unit startup with NC system at normal operating pressure and temperature would pose a hardship for the station because of the high risk of an inadvertent pressurizer auxiliary spray initiation to the pressurizer at normal NC system operating temperature and pressure. This "Upset Condition" design transient is undesirable for two reasons:

1)

It would force static piping "cold water" contents into the pressurizer spray line and result in an additional thermal design cycle. The plant design only allows for 10 of these over the plant design life (ref. Updated Final Safety Analysis Report (UFSAR), Table 5.2, "Summary of Reactor Coolant System Design Transients").

2)

It would violate the maximum differential between spray water injected and the pressurizer water temperature allowed by the Selected Licensee Commitment Manual (UFSAR Chapter 16). Section 16.5.8.c requires an engineering evaluation to determine the effects of the cold water on the structural integrity of the pressurizer.

- 6 Opening NO-1 during RCS pressurization to pressurize the portion 2 piping would pose a hardship for the station because it would violate 10 CFR 50.55a(c)(ii) that requires a double isolation valve barrier of the RCS boundary from the NO system. This would create an inability to mitigate a loss of coolant accident (LOCA) if a break was to occur in the 14-inch piping between valves NO-1 and NO-2, reducing the plant's margin of safety. Valve NO-1 could not be counted on to close against the postulated flow from the RCS through a 14-inch line break. It would also subject NO system components to risk of damage with only a single valve isolation from RCS pressure.

Also, with valve NO-1 open, the :X-inch vent line and test header branch line (through valve NO

91) would be required to serve as part of the double isolation valve RCS boundary. While the

X-inch piping between valves NO-54, NO-90, and NO-91 is designed for RCS pressure and temperature, two issues of concern are:

1.

A breach or break in either of the :X-inch lines between valves NO-54, NO 90, and NO-91 in this alignment would challenge the unit's stability and result in an unplanned unit shutdown for piping/component repair.

2.

Any pre-existing minor leakage (non-problematic in the "normal" startup valve alignment) past valve 1NO-54 would likely be intensified by direct, unobstructed RCS pressure applied to the valve's seat, and pressurize the short section of non ASME Code (Duke Piping Class E) :X-piping between the three valves to RCS pressure. This would charge valves 1NO-90 and 1NO-91 (and their seats) with RCS pressure, risking an RCS pressure boundary leak as well as possible valve seat damage.

On the portion 3 piping, no intermediate test connection exists on the 3-inch segment of pipe between these check valve pairs to measure the test pressure locally. Aligning an NV pump to the boron injection flow path in Mode 3 (at startup) and cracking open valve NI-3 would constitute a manual safety injection, counting against the allowed cold leg thermal design transients (design limit is 50 for the life of the plant). Such action would pose a hardship for the plant.

Installing temporary jumpers on portion 4 or portion 5 piping to transfer RCS pressure between the pressure isolation valve (PIV) check valves during startup, with NC system at RCS pressure, would be done in Mode 3. At this point, the RCS PIVs have already undergone PIV leak rate testing per Technical Specification (TS) Surveillance Requirement (SR) 3.4.14.1 to verify their leakage is within TS limits. Introducing RCS pressure between the PIV check valves at this time would likely cause the inboard PIV check valve to unseat, placing the station in TS 3.4.14 LCO Action Condition.

The PIV's serve as the RCS pressure boundary (see 10 CFR 50.2 and 10 CFR 50.55a(c)). The limit on allowable PIV leakage rate prevents over-pressurization to the low-pressure portions of the connecting NI system piping as well as the loss of integrity of a fission product barrier.

Failure consequences could be a LOCA outside of containment, which is an unanalyzed accident that could degrade the ability of low pressure injection.

- 7 Using a hydro pump to pressurize portion 4 or portion 5 piping during no mode would in either case, require all 4 of the inboard NI check valves interfacing the NC system (1 a-inch for portion 4 and 6-inch for portion 5) to be temporarily gagged closed to provide a pressurization boundary and would be a hardship for the station.

RCS pressure could be applied to the portion 6 and/or portion 7 piping by opening the "inside" NC loop isolation valves (including "in series" valves NC-19 and NC-20 for loop 4) at the onset of NC system pressurization. Each of these valves is the first of a series of two valves maintaining double-isolation of the RCS pressure boundary either from other piping or from the containment atmosphere.

Opening these valves to pressure test this piping at RCS pressure would eliminate the double valve protection required by 10 CFR 50.55a(c)(ii) for the RCS boundary, creating a "single valve barrier" between the RCS pressure boundary and non-code piping, some of which is vented to the reactor containment atmosphere. The piping on the discharge side of these "single valve barriers" is non-Code piping, and not designed to serve as part of the RCS double isolation pressure boundary.

Additionally, Selected Licensee Commitment 16.5.10 prohibits opening one of the portion 7 reactor head vent valves with the unit in Modes 1 through 4.

Since no isolation is possible from either the reactor vessel or the pressurizer relief tank without significant modification, and no connections exist between the valve pairs for test connection, testing of the piping between the reactor head vent pipe double isolation valves by hydro pump or temporary jumper is not possible.

The licensee believes that any increase in confidence on piping integrity attained by pressurizing to 2235 psig of the portions 1 through 7 would not be commensurate with the increase in radiation exposure and/or safety hazards that the station personnel would be subjected to.

3.1.5 Licensee's Proposed Alternative Through-wall leakage that can be detected at a given pressure such as at the ReS pressure can also be detected at a lower pressure with reduced leakage rate depending on the pressure ratio. It may take longer for some leaks to propagate through the piping wall at lower pressures, but generally, during reduced pressure testing, the resulting leak rates would be reduced, but the leakage would still be visible to VT-2 examination.

It has been established that leakage through a fixed area orifice varies proportional to the square root of the ratio of the differential pressures as stated below. For example, if a leak L were projected to be present at 2235 psig, that same leak would be present at 327 psig, but with a magnitude of:

~327;2235 x L =0.38L Inspections that reveal no leakage at 300 psig or above (where 38% of the leakage produced by 2235 psig pressures would be present for detection during VT-2 examination), therefore, give

- 8 reasonable confidence that no leakage would be present at 2235 psig. The pressure used in the reduced pressure testing performed as alternative pressure test covered in this RR is at least 300 psig or higher if system condition allows at the time of testing.

The pressure tests performed at the lower pressures stated in the proposed alternative pressure tests provide reasonable assurance of structural integrity. Pursuant to the 10 CFR 50.55a(a)(3)(ji), compliance to the Code requirement would result in hardship without a compensating increase in the level of quality and safety 3.2 NRC Staff's Evaluation The NRC staff has evaluated hardship to the licensee in performing the Code-required system leakage test of those components listed in the licensee's submittal, at a test pressure corresponding to the pressure associated with 100% of rated reactor power. The components in each case are connected to the RCS and are normally isolated from direct RCS pressure during normal operation. They are isolated from the reactor coolant loop by their location, either upstream of a check valve, between two check valves, or between two closed valves that must remain closed during the unit's operation (or startup) in Modes 3,2, or 1 The NRC staff believes that application of the RCS pressure to the Portion 1 piping could result in the actuation of auxiliary sprays into the pressurizer and, thus, has the potential to initiate a design transient which would affect the design life of the plant.

Application of the RCS pressure to the portion 2 piping during a system pressure test would appear to not be in conformance with the provisions of 10 CFR 50.55a(c)(2)(ii) which require a double valve isolation barrier between the RCS and the RHR system. With only one isolation valve separating the RCS from the RHR system, the system will not mitigate a LOCA if a break were to occur in between the boundary valves or preclude pressurizing the low pressure RHR system to higher than its design pressure due to an interface system LOCA. Applying RCS pressure to portions 6 and 7 would require valves to be opened that would remove the double valve protection required by 10 CFR 50.55a(c)(2)(ii) and leave only a single valve as the barrier between the RCS and the non-Code piping.

The portion 3 piping has no provision to connect an external pump to test the piping segment, and in order to pressurize the portion, one of the charging pumps must be aligned that would inject cold borated water into the RCS with the potential to cause a cold leg thermal design transient and, thus, affect the plant's design life.

For portions 4 and 5 to be pressurized to the RCS pressure, temporary jumpers need to be installed between the RCS and the subject piping. Alternatively, a separate hydro pump could be used that would require gagging of NI check valves to hold the RCS pressure. Gagging would cause increased loads on the valve bodies and seats to hold them shut, thus leading to an increased risk of damage to the valves.

The NRC staff has reviewed the issues related to performing system leakage tests at the RCS pressure corresponding to 100% rated reactor power of 2235 psig and has evaluated the licensee's proposed alternative test pressure of at least 300 psig or the system operating pressure at the time of testing.

- 9 The NRC staff has further determined that there are hardships associated with the test being performed in accordance with the referenced Code due to the following:

Special valve line-ups for these tests add unnecessary challenges to maintaining system configuration. Tests performed inside the containment :increase radiation exposure to plant personnel during modification, restoration of system line-ups and the removal of contaminated test equipment.

Components are exposed to test conditions above their normal operating pressure and temperature supported by temporary connections.

Use of single valve isolation is also a significant personnel safety hazard.

The NRC staff believes that use of a higher testing pressure for components that would not normally be subject to RCS pressure during 100% rated reactor power would not result in an increase in assurance of structural integrity that would be commensurate with the potential challenges to the facility and additional radiation exposure to plant personnel.

A mitigating factor in accepting the test pressure at or above 300 psig in lieu of the ASME Code required test pressure is based on the fact that there is no known degradation mechanism, such as intergranular stress corrosion cracking primary water stress corrosion cracking, or thermal fatigue that is likely to affect the welds in the subject segments.

Additionally, each portion of piping listed in licensee's relief request will get a VT-2 visual examination subsequent to alternative testing described in the RR with the RCS at full temperature and pressure and in its normal alignment, as part of the inspection boundary for the 1O-year interval Class 1 system leakage test performed during unit start-up in Mode 3.

In an unlikely event of not detecting a leak in the subject piping segments, the instrumentation available to the operators for detection and monitoring of RCS leakage would provide prompt qualitative information to permit them to take immediate corrective action. Therefore, the NRC staff has determined that compliance with the ASME Code requirements stated in paragraph IWB-5221 and in regard to the required test pressure during system leakage test for the components identified in the licensee's RR would result in hardship without a compensating increase in the level of quality and safety.

5.0 CONCLUSION

The NRC staff concludes that for components identified in RR 09-MN-005, performance of the system leakage test conducted at a test pressure corresponding to the operating pressure of the component stated in the licensee's proposed alternative, in lieu of testing at the RCS pressure associated with 100% rated reactor power as required by the Code, is acceptable. The NRC staff has determined that for this portion of the Class 1 boundary, compliance with the requirement of the Code of Record as stated in paragraph IWB-5221 in regard to the test pressure during the system leakage test would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety and, therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee's proposed alternative is authorized for McGuire 1 and 2 for the remainder of the third 1O-year inspection interval.

- 10 All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: P. Patnaik, NRR Date: June 14, 2010

ML101580422

• concurrence via memo dated 4/9/10 ML100990009 OFFICE NRRlLPL2-1/PM NRR/LPL2-1/LA NRR/CPNB/BC NRR/LPL2-1/BC NRR/LPL2-1/PM NAME JThompson MO'Brien RTaylor' GKulesa (VSreenivas for)

JThompson DATE 06/09/10 06/09/10 04/09/10 06/14/10 06/09/10