ML120820510

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Issuance of Amendment Revision to the Applicability of the Reactor Coolant System Pressure-Temperature Limits and the Cold Overpressure Protection Setpoints
ML120820510
Person / Time
Site: Seabrook 
Issue date: 04/15/2013
From: John Lamb
Plant Licensing Branch 1
To: Walsh K
NextEra Energy Seabrook
Lamb J, NRR/DORL, 301-415-3100
References
TAC ME7645
Download: ML120820510 (22)
v • d • e


Text

{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 15, 2013 Mr. Kevin Walsh Site Vice President c/o Michael O'Keefe Seabrook Station NextEra Energy Seabrook, LLC P.O. Box 300 Seabrook, NH 03874

SUBJECT:

SEABROOK STATION, UNIT NO.1-ISSUANCE OF AMENDMENT RE: REVISION TO THE APPLICABILITY OF THE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS AND THE COLD OVERPRESSURE PROTECTION SETPOINTS (TAC NO. ME7645)

Dear Mr. Walsh:

The Commission has issued the enclosed Amendment No. 135 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1 (SeabrOOK). This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 17, 2011, as supplemented December 3, 2012, and January 9, 2013. The amendment revises the applicability of the figures in the TSs for the reactor coolant system pressure-temperature limits and the cold overpressure protection setpoints. The amendment revises the applicability of the figures from 20 effective full-power years (EFPY) to 23.7 EFPY. A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. John . Lamb, Senior Project Manager Plant icensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443

Enclosures:

1. Amendment No. 135 to NPF-86
2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 NEXTERA ENERGY SEABROOK, LLC, ET AL.* DOCKET NO. 50-443 SEABR9_QK.$TATION1...illi!T NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 135 License No. NPF-86

1.

The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment filed by NextEra Energy Seabrook, LLC, et aI., (the licensee) dated November 17,2011, as supplemented December 4, 2012, and January 9, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (tne Act), and the Commission's rules and regulations set forth in 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: 0) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and Oi) 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 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.

  • NextEra Energy Seabrook, LLC is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Light Plant and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

- 2

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 Facility Operating License No. NPF-86 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 135, and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days. FOR THE NUCLEAR REGULATORY COMMISSION Meena K. Khanna, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: April 15, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 135 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change. Remove 3 Replace the following pages of the Appendix A. Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Insert vi vi 3/44-23 3/44-23 3/44-24 3/44-24 3/44-30 3/44-30

- 3 (4) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, 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; (5) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, 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 calibration or associated with radioactive apparatus or components; (6) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30,40, and 70, to possess, but not separate, such byproduct and speCial nuclear materials as may be produced by the operation of the facility authorized herein; and (7) DELETED C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable prOVisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power). (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 135", and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook. LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) License Transfer to FPL Energy Seabrook. LLC**

a.

On the closing date(s} of the transfer of any ownership interests in Seabrook Station covered by the Order approving the transfer, FPL Energy Seabrook, LLC**, shall obtain from each respective transferring owner ali of the accumulated decommissioning trust funds for the facility, and ensure the deposit of such funds and additional funds, if necessary, into a decommissioning trust or trusts for Seabrook Station established by FPL Energy Seabrook, LLC", such that the amount of such funds deposited meets or exceeds the amount required under 10 CFR 50.75 with respect to the interest in Seabrook Station FPL Energy Seabrook, LLC**, acquires on such dates(s).

  • Implemented
    • On April 16, 2009, the name "FPL Energy Seabrook, LLC" was changed to "NextEra Energy Seabrook, LLC".

AMENDMENT NO. 13 5

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION FIGURE 3.4-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY> 1jJCi/gram DOSE EQUIVALENT 1-131.......................................................................... 3/44-20 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................................................................................. 3/44-21 3/4.4.9 PRESSUREITEMPERATURE LIMITS General............................................................................................................ 3/4 4-22 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LlMITATIONS-APPLICABLE UP TO 23.7 EFPY................................................................ 3/44-23 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LlMITATIONS-PPLICABLE UP TO 23.7 EFPY................................................................... 3/44-24 Pressurizer...................................................................................................... 3/4 4-25 Overpressure Protection Systems............................................................... 3/4 4-26 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS................ 3/44-30 3/4.4.10 DELETED.................................................................................................................... 3/44-31 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................................................... 3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup. and Power Operation.......................................................... 3/45-1 Shutdown..................................................................................................................... 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350°F............ 3/45-4 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350°F.................................................. 3/4 5-8 ECCS SYBSYSTEMS - Tavg Equal To or Less Than 2000F................................ 3/45-10 3/4.5.4 REFUELING WATER STORAGE TANK................................................................ 3/45-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity................................................................................................. 3/46-1 Containment Leakage................................................................................................ 3/4 6-2 SEABROOK - UNIT 1 vi Amendment No. 70. 89,115, 126, 135 Corrected By Letter Dated May 17, 2007

MATERIAL PROPERTY BASIS Limiting material: LOWER SHELL PLATE R-1S00-1 limiting ART values at 23,7 EFPY:1/4T, 109'F 314T, SS'F Curves applicable for the first 23.7 EFPY and contain margins of 20'F and 100 psig for possible Instrument errors 2eoo~~~~~--~~~~~~~~~~~I~~~--'~~~~~~~~~~ t I I I. til I I ...J~ _1 _ _ I ~.J~ ~1...... _1 __ 1 ~ _1 __ ~ _.1 __,_ 1_..1_.. I... J __ 1__ l_-I,.._f._.1 _'- _ L _..I __ 1-_,J.. _f__ I I' j I I I,! I I I I I I I f I I t ' t i l \\ I I I I I I I _~-~~~T_~_-~-1--I-w.-~--~-~--'--~.1~-~~1-*~-.-~--r-' -.-~-~~-1*-~~ 2600 I f I I f ' I C I I I I I f I I t I I 1 I I ~ I I I I j I I I I I ,... -,"".. r" -.- -,.., - -,- T'" "I"' -{" -, - -C.. r - "'11-.. r.., -,- - r - - r.. 1.. -I"".. r... "'," - r...,.. -,- I I I t I I I I I t I Itt I I I I I i j .l I I I .. J.. 1.....J... _L _J.....1_ _ L...J.... L.. J..... *_ '" J. _'... _L _ J.....'.,. _l __' __ L.....! _... 1__ 2400 Leak Test Limit I I I I I . i.... -I-I _.l _ _ I_ I Unacceptable Operation Heatup Rate 80 Deg. FlHr ..1__, _...I..... 1- _....... i. i I I I I ) I I,, . - 'i I Heatup Rate _J I ( I I I I I 100 Deg. F/Hr I 1 r I I I I fit ... 01... _1 __ ~ _.... _.. \\. _ -I.... _1_...... _ I I ..,.. -,-... T"., - - r..,.. ""I'"... T _~~~.1 .. '_",.l. __I~ I j I I I I 1 ~":- ~"_.: - -;- ~; - ~;.. - -:-.. ~.. ~--:-.. r.. -:-~I 1 I I I J __ I... _......; __

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VAUD FOR THE FIRST 23.7 EF?Y, SETPOtNT CONTAINS MARGIN OF SOQF r-oR TRANSIENT EFFECTS T s 200.0oF I ? = 561.0 PSIGi

ZOO.O°F < T ~ 230.SDf I P = 12.1*(T-200.0) + 926.0 PSIGi 230.SoF < T $ 255.0oF t P "" 23.15*(T-230.S) + 1295.05 PSIGi T> 2S5.0oF I P ::: 34.5*(T-255.0) + 1B62.225 PSIG 2500~------~------~------~------~--~--~---------~

2250 2000 1''''''''''''''''''''''''''''''''''''''''''' ~ 1750 ' Il~ ~ 1250 a:: ~ x: 1000 ' I 750 500 O--------~------~--~----------~------~------- 50 100 150 200 250 300 350 RCS TEMPERATURE (DEG. F) FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS SEABROOK - UNIT 1 3/44-30 Amendment No. 89,115,116, 135

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.135 TO FACILITY OPERATING LICENSE NO. NPF-86 SEABROOK STATION, UNIT NO.1 DOCKET NO. 50-443

1.0 INTRODUCTION

By letter dated November 17, 2011, as supplemented December 12, 2012, and January 9, 2013,1 NextEra Energy Seabrook, LLC (NextEra or the licensee) submitted license amendment request (LAR) LAR 11-06 to revise the applicability of the figures in the Technical SpeCifications (TSs) for the reactor coolant system (RCS) pressure-temperature (P-T) limits and the cold overpressure protection system (COPS) setpoints at Seabrook Station, Unit No.1 (Seabrook). The proposed amendment revised the applicability of the TS figures from 20 effective full-power years (EFPY) to 23.7 EFPY, based on a revised neutron fluence evaluation. The proposed change to the applicability term would not affect the actual P-T limit curves and COPS setpoints, which were originally approved for 20 EFPY in License Amendment No. 89.2 The supplements dated December 12, 2012,and January 9, 2013, provided additional information that clarified the application, did not expand the; scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on January 10, 2012 (77 FR 1519).

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC) has established requirements in Appendix G, "Fracture Toughness Requirements," of Part 50 to Title 10 of the Code of Federal Regulations (10 CFR Part 50) in order to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. As required in 10 CFR Part 50, Appendix G, the P-T limits for an operating light-water nuclear reactor must be at least as conservative as those that would be generated if the methods of Appendix G, "Fracture Toughness Crit.eria for Protection Against Failure," to Section Xl of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the ASME Code) were used to generate the P-T limits. Additionally, as required in 10 CFR Part 50, Appendix G, applicable surveillance data from reactor pressure vessel (RPV) material surveillance programs must be incorporated into the calculations of plant specific P-T limits and the P-T limits for operating reactors must be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials. 1 Agencywide Documents Access and Management System (ACAMS) Accession Nos. ML11329A017, ML12341A095 and ML13014A624, respectively. 2 ADAMS Accession No. ML032250621.

- 2 Table 1 of 10 CFR Part 50, Appendix G provides the NRC staff's criteria for meeting the P-T limit requirements of the ASME Code, Section XI, Appendix G, as well as the minimum temperature requirements of the rule during normal and pressure testing operations. In addition, the NRC staff's regulatory guidance related to the evaluation of neutron embrittlement for P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials." Additional guidance related to the staffs review of P-T limit curve submittals is found in NUREG-0800, Standard Review Plan Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock." The ASME Code, Section XI, Appendix G methodology for generating P-T limit curves is based upon the principles of linear elastic fracture mechanics (LEFM). The basic parameter of this methodology is the stress intensity factor, KI, which is a function of the stress state in the component and flaw configuration. The ASME Code, Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic and pressure testing limits. The ASME Code, Section XI, Appendix G specifies that the P-T limits shall be generated by postulating a flaw with a depth that is equal to 1/4 of the RPV section thickness and a length equal to 1.5 times the RPV section thickness. The critical locations in the RPV section thickness for calculating heat-up and cool-down P-T limit curves are the 1/4 thickness (1/4T) and 314 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively. P-T limit curve calculations are based, in part, on the reference nil-ductility temperature (RTNDT) for the material, as specified in the ASME Code, Section XI, Appendix G. The RTNDT is the critical parameter for determining the critical or reference stress intensity factor (fracture toughness, Klc) for the material. As required by 10 CFR Part 50, Appendix G, RTNDTvalues for materials in the RPV beltline region shall be adjusted to account for the effects of neutron radiation. Regulatory Guide 1.99, Rev. 2, contains methodologies for calculating the adjusted RTNDT (ART) due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the shift in reference temperature caused by irradiation (8RTNDT), and a margin term. The 8RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper (Cu) and nickel (Ni) in the material and may be determined from the tables in RG 1.99, Rev. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the postulated flaw depths described above. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the neutron fluence and the calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term. To satisfy the requirements of 10 CFR Part 50, Appendix G, methods for determining fast neutron l'luence are necessary to estimate the fracture toughness of the RPV materials. Appendix H, "Reactor Vessel Material Surveillance Program Requirements," of 10 CFR Part 50, requires the installation of surveillance capsules, including material test specimens and flux dosimeters, to provide data for material damage correlations as a function of neutron fluence. In March 2001, the NRC staff issued RG 1.190, "Calculational and DOSimetry Methods for Determining Pressure Vessel Neutron Fluence." Regulatory Guide 1.190 describes methods

- 3 and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to meeting the regulatory requirements discussed above.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation The licensee's LAR dated November 17, 2011, proposes the following changes to the TS figures for the P-T limit curves and COPS setpoints:

  • The title of TS Figure 3.4-2, "Reactor Coolant System Heatup Limitations - Applicable Up to 20 EFPY," is revised to specify an applicability term of 23.7 EFPY.
  • The title of TS Figure 3.4-3, "Reactor Coolant System Cooldown Limitations - Applicable Up to 20 EFPY," is revised to specify an applicability term of 23.7 EFPY.
  • The title of TS Figure 3.4-4, "RCS Cold Overpressure Protection Setpoints Valid for the First 20 EFPY, Setpoint Contains Margin of 50 OF for Transient Effects," is revised to specify an applicability term of 23.7 EFPY.

Additionally, for all other instances where the value of 20 EFPY appears on the TS figures above, it is replaced with 23.7 EFPY. The TS index is similarly revised to reflect the change in the applicability term, consistent with TS Figures 3.4-2, 3.4-3, and 3.4-4. The licensee stated that Seabrook had accumulated 17.85 EFPY at the end of Cycle 14 in the Spring of 2011. Therefore, the TS P-T limit applicability period of 20 EFPY is projected to be reached during Cycle 16, and revised TS figures are required to support operation beyond Cycle 16. According to the licensee, methodologies employed for developing the current TS P-T limit curves for 20 EFPY are documented in WCAP-15745, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," December 2001, which was submitted to the NRC in an October 2002 LAR3 to implement the curves. The 20 EFPY P-T limit curves and COPS setpoints were approved by the NRC in License Amendment No. 89, issued by letter dated September 11, 2003. The licensee stated that the 20 EFPY P-T limits incorporated the analysis of neutron dOSimetry for surveillance capsules U and Y and were developed based on the latest neutron fluence projections available at the time. The 20 EFPY neutron fluence projections were based on actual core power distributions for Cycles 1 through 5 and then projected forward to 20 EFPY. The licensee's LAR submittal included Table 5 from WCAP-15745, which shows the 20 EFPY neutron fluence projections for key azimuthal locations at the RPV clad/base metal interface. The licensee indicated that surveillance capsule V was removed from the RPV at the end of Cycle 10 and tested. The test results for capsule V are documented in WCAP-16526-NP, "Analysis of Capsule V from FPL Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program," March 2006,4 which was submitted to the NRC in a letter dated April 10, 3 ADAMS Accession No. ML022940024. 4 ADAMS Accession No. ML061030088.

- 4 2006. 5 The licensee stated that the analysis of capsule V dosimetry indicated that the 20 EFPY applicability period for the P-T limit curves is conservative. To develop new P-T limit curves for future LAR submittals, Westinghouse developed WCAP 17441-NP, "Seabrook Unit 1 Heatup and Cooldown Curves for Normal Operation," October 2011.6 WCAP-17441-NP documents P-T limit curve calculations for 36 EFPY and 55 EFPY. The licensee used the revised neutron fluence evaluation from WCAP-17441-NP as the basis for extending the P-T limit curve applicability period to 23.7 EFPY. It should be noted that WCAP-17441-NP was not included with the subject LAR. However, the LAR did provide Table 2-3 from the WCAP, which lists a revised set of neutron fluence projections out to 60 EFPY, based on cycle-specific calculations for Cycles 1 through 14. The licensee noted that fluence projections beyond Cycle 14 incorporated a 13 percent uncertainty factor based upon the guidance in RG 1.190, The licensee stated that a comparison of the neutron fluence projections from WCAP-15745 with the revised fluence projections from WCAP-17441-NP indicates that, accounting for the data retrieved from the additional operating cycles since the WCAP-15745 neutron fluence calculations, the actual rate of neutron fluence accumulation is lower than that originally considered in the development of the current P*T limit curves. Considering this conservatism, Westinghouse performed an applicability evaluation of the current TS Figures 3.4-2, 3.4-3, and 3.4-4, which justifies the extension of the applicability period to 23.7 EFPY, based on the updated fluence evaluation from WCAP-17441-NP. The applicability evaluation is provided in an enclosure to the LAR. The licensee noted that the limiting ART values used for developing the P-T limit curves were based upon the currently approved 20 EFPY projected RPV neutron fluence. Since the LAR's applicability evaluation shows that this same neutron fluence projection is reached at 23.7 EFPY, the limiting ART values are not impacted. The licensee further noted that the COPS setpoints are based on the P*T limits and are likewise not impacted. Therefore, the licensee concluded that, based upon the updated neutron fluence and applicability evaluation, the change in applicability term is justified, and the present P-T limits and COPS setpoints may be implemented until 23.7 EFPY with no impact on safety margin. 3,2

NRC Staff Evaluation

The licensee's LAR submittal proposes to revise the applicability period for the current TS P-T limit curves from 20 EFPY to 23.7 EFPY of facility operation. The revision to the applicability period is based on a revised neutron fluence assessment, wherein the licensee determined that the projected neutron fluence values used as the basis for calculating the limiting beltline shell material ART values and P-T limit curves would be reached at 23.7 EFPY, rather than 20 EFPY. The actual TS P-T limit curves, including the limiting ART values and corresponding neutron fluence inputs on which they are based, are not affected by the proposed TS change. The low temperature overpressure protection function at Seabrook is provided by the COPS, which is enabled at a predetermined temperature to prevent RCS pressure from exceeding the p.T limits at low temperatures. The LAR also proposes to extend the applicability of the TS 5 ADAMS Accession No. ML061030087. 6 ADAMS Accession No. ML12341A096.

- 5 COPS setpoints from 20 EFPY to 23.7 EFPY, based on the results of the revised neutron fluence assessment. The COPS pressure setpoints are established to ensure RCS operation within the required P-T limits and are calculated based on the consideration of design-basis heat-addition and mass-addition transients. Since the P-T limit curves, limiting ART values, and design-basis transients are not impacted by the proposed TS change, the COPS arming temperature and pressure setpoint calculations are also not affected. Therefore, the NRC staff's detailed review of the subject LAR is confined to evaluating the acceptability of the p.T limit curves for 23.7 EFPY. 3.2.1 limiting ART Values and P-T limit Curves The Seabrook TS P-T limit curves were originally established for 20 EFPY based on the calculations documented in Westinghouse report WCAP-15745, which was included with the LAR dated October 11, 2002, to implement the curves. The Seabrook COPS setpoints were established for 20 EFPY using the methodologies documented in Framatome ANP report NFSP 02-0061, "Seabrook Station Cold Overpressure Mitigating System (COMS) Setpoint Development Methodology," dated August 2002, which was also provided with the LAR dated October 11, 2002. By letter dated September 11, 2003, the NRC staff issued Seabrook license Amendment No. 897 approving the implementation of the current TS P-T limit curves and COPS setpoints for 20 EFPY. The Seabrook TS P-T limit curves were generated based on the analysis of the limiting beltline she" material, Lower She" Plate R-1808-1 (Heat Identification No. D1081-3). As discussed in its September 11, 2003, safety evaluation (SE) accompanying license Amendment No. 89, the NRC staff independently verified that the ART values for this material at the 1/4T and 3/4T locations were correctly calculated using the procedures in RG 1.99, Rev. 2, Position 1.1 using valid input parameters, as summarized in Table 1 below: Table 1 - Calculation of limiting ART Values at the 1/4T and 3/4T Locations for 20 EFP I Material Lower Shell Plate R-1801-1 Location 1/4T i Initial RTNDT (OF) 40 Fluence at RPV Clad/Base Metal Interface (n/cm2) 1.324 X 1019 i Fluence at Location (n/cm2) 7.89 X 1018 CF (OF) 37 (table(1>>) ~ (OF) 34.6 Margin,L} (OF) 34 ART (OF) 109 Lower Shell Plate R-1801-1 3/4T 40 i 1.324 X 1019 i 2.90 x 1018 I 37

(table(l>>).

24.2 24.2 i 88 I i (1) CFs were determined using Table 2 from RG 1.99, Rev. 2, as prescribed in Position 1.1 of the RG. (2) The margin term for each ART calculation was based on the establishment of an initial material property uncertainty (01) and shift in material property uncertainty (01-:'), consistent with the guidance in RG 1.99, Rev. 2. 7 ADAMS Accession No. ML032250621.

- 6 As discussed in its SE dated September 11, 2003, the NRC staff performed an independent calculation to verify that the Seabrook P-T limit curves generated in WCAP-15745 are at least as conservative as those that would be generated using the methods and acceptance criteria of the ASME Code, Section XI, Appendix G, as required by 10 CFR Part 50, Appendix G. The NRC staff had also confirmed that the P-T limit curves included appropriate minimum temperature requirements that are at least as conservative as those required by Table 1 of 10 CFR Part 50, Appendix G. 3.2.2 Revised RPV Neutron Fluence Evaluation The 20 EFPY neutron fluence projection was originally calculated in WCAP-15745 based on actual core power distributions for Cycles 1 through 5 and the analysis of surveillance capsule U and Y neutron dosimetry. Updated neutron fluence projections prepared by Westinghouse in WCAP-17441-NP, based on actual core thermal power histories through Cycle 14, determined that the 20 EFPY applicability period for the current curves is overly conservative. The LAR provided Table 2-3 from WCAP-17441-NP, which lists cycle-specific neutron fluence values at the RPV clad/base metal interface for Cycles 1 through 14 and neutron fluence projections through 60 EFPY. A comparison of the neutron fluence projections from WCAP-15745, and the projections based upon actual Cycle 1 through 14 data shown in WCAP-17441-NP, Table 2-3, indicates that, accounting for the data retrieved for the additional operating cycles since the previous fluence calculations, the actual rate of fluence accumulation is lower than that originally considered for developing the curves. Accordingly, the updated neutron fluence assessment from WCAP-17441-NP was used as the basis for extending the applicability of the current TS P-T limit curves from 20 EFPY to 23.7 EFPY. It should be noted that WCAP-17 441-NP was prepared to support future P-T limit curve submittals for 36 EFPY and 55 EFPY and was not submitted for staff review with the LAR; only the revised neutron fluence values from Table 2-3 of the WCAP were provided in the LAR submittal. The NRC staff noted that the portions of WCAP-17 441-NP pertaining to neutron fluence must be reviewed in order to determine whether the methods used for calculating the revised neutron fluence values are adherent to RG 1.190. Therefore, in RAI-1 the NRC staff requested that the licensee provide WCAP-17441-NP, or those portions of WCAP-17441-NP pertaining to neutron fluence. By letter dated December 3,20128 the licensee provided WCAP-17441-NP. The NRC staff reviewed the WCAP report and confirmed that the neutron fluence prOjections were calculated using the methods described in WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Rev. 4, May 2004. 9 As documented in the NRC SE approving WCAP-14040-A, Rev. 4, these neutron fluence calculational methods adhere to the guidance contained in RG 1.190. Based on this consideration, the NRC staff determined that the fluence calculations described in WCAP-17441-NP are acceptable. The licensee's EFPY applicability evaluation was based on a linear interpolation between the revised neutron fluence values of WCAP-17 441-NP to determine that the TS P-T limit curves WOUld, in fact, be valid for 23.7 EFPY. The NRC staff performed an independent calculation, 8 ADAMS Accession No. ML12341A095. 9 ADAMS Accession No, ML050120209,

- 7 which verified that the 23.7 EFPY operating term corresponds to the peak RPV neutron fluence of 1.324 x 1019 n/cm2 (E > 1.0 MeV), based on a linear interpolation between the revised neutron fluence values. Therefore, the NRC staff finds that the licensee's P-T limit curves for the limiting beltline shell material, Lower Shell Plate R-1808-1, are acceptable for 23.7 EFPY. 3.2.3 Consideration of Ferritic RCPB Components Outside of the Immediate RPV Beltline Shell Region Regarding ferritic RCPB components that are not part of the RPV belt/ine shell region, 10 CFR Part 50, Appendix G, and Paragraph IV.A states the following: The pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials must meet the requirements of the [ASME Code, Section III], supplemented by the additional requirements set forth in [paragraph IV.A.2, "Pressure-Temperature (P-T) limits and Minimum Temperature Requirements"]... Therefore, 10 CFR Part 50, Appendix G requires that P-T limits be developed by considering beltline and non-beltline ferritic RCPB components. Further, 10 CFR Part 50, Appendix G requires that all ferritic RCPB components must meet the applicable ASME Code, Section III requirements. For ferritic RCPB piping, pumps and valves greater than 2.5 inches in thickness, the relevant ASME Code, Section III requirement that will affect the P-T limits is the lowest service temperature (LST) requirement specified in Section III, NB-2332{b). The NRC staff noted that P-T limit calculations for ferritic RCPB components that are not RPV beltline shell materials may define P-T limits that are more restrictive than those calculated for the RPV beltline shell materials. This may be due to the following factors: (1) RPV nozzles, penetrations, and other discontinuities exhibit higher stresses than the RPV beltline shell region, which could result in more restrictive P-T limits, even if the RTNDTfor these components is not as high as that of the RPV beltline shell materials. (2) Ferritic RCPB components that are not part of the RPV may have initial RTNDTvalues, which may define a more restrictive LST in the P-T limits than the minimum temperature requirements for the RPV. Therefore, in RAI_2,10 the NRC staff requested that the licensee describe how the P-T limit curves, and the methodology used to develop these curves, considered all ferritic RCPB components, consistent with the requirements of 10 CFR Part 50, Appendix G. In its response dated January 9,2013, to RAI-2, factor (1), the licensee stated that the 20 EFPY TS P-T limits were developed using the generic methodologies described in Topical Report WCAP-14040-A, Rev. 2, January 1996. The licensee noted that WCAP-14040, Rev. 2 did not consider the inlet/outlet nozzles, which are the most highly stressed ferritic component of the RPV. Therefore, to demonstrate that the P-T limit curves are bounding for the entire RPV, the licensee developed component-specific P-T limit curves for the RPV inlet/outlet nozzles. The 10 ADAMS Accession No. ML13014A624.

- 8 NRC staff's evaluation of the licensee's response to factor (1), including its P-T limit curve calculations for the inlet/out nozzles, is provided below. The licensee provided ART calculations for the nozzle forgings and the associated nozzle-to shell welds based on 55 EFPY RPV nozzle fluence values from WCAP-17441-NP. The licensee determined that the 55 EFPY neutron fluence values are 1.02 x 1017 n/cm2 (E > 1.0 MeV) at the lowest extent of the inlet nozzles and 5.70 x 1016 n/cm2 (E > 1.0 MeV) at the lowest extent of the outlet nozzles. The licensee noted that these 55 EFPY neutron fluence values are conservative and bounding relative to the consideration of these nozzles for the 23.7 EFPY P-T limit curves. The 55 EFPY ART calculations for the Seabrook inlet/outlet nozzle forgings and associated nozzle-to-shell welds, along with the critical input parameters (55 EFPY fluence, Cu, Ni, and initial RTNDTvalues), are shown below. Table 2 - Calculation of ART Values for the Inlet/Outlet Nozzle Forgings and Nozzle-to-Shell Welds at the 1/4T(1) Location for 55 EFPY Material Cu I Ni ,CF from I Surface Fluence at .1RTNOT Margin ART Content! Content! RG 1.99. Fluence 1/4T ~ (OF) (OF) (Weight (Weight* Tables (n/cm2) Location %) %) (OF) (n/cm2) Inlet Nozzles 0,10 67 1,02 X 5,62 x 0.89 4.91 4.90 9.8 o 1017 16 Bounding 10 Case(2) I Nozzles Outlet -10 0.67 240 5.70 X 3.31 X 11.80 13.5 OAO 11.80 16 16 Bounding 10 10 Case(2) Inlet Nozzle to Upper 1.02 X 0.35 1.00 272 6.10 X 21 40 5.0 1017 1016 Shell Welds Outlet Nozzle to Upper -56(3) 0.35 1.00 272 5.70 x 3,40 x 13,60 36,60 -5.8 16 1016 Shell i 10 i Welds I i i I (1) Only the 1/4T location was analyzed for the nozzle evaluation, as discussed below, (2) Seabrook, a Westinghouse 4-loop plant, has four inlet nozzles and four outlet nozzles. The licensee selected the most limiting of the four measured values for initial RTNOT, Cu content, and Ni, content. (3) The licensee used a generic initial RTNoTvalue of -56 OF for these Linde 0091 submerged arc welds, and the margin term was calculated using a 0'1 value of 17 OF. This is consistent with 10 CFR 50.61 and RG 1,99, Rev, 2, The NRC staff confirmed that the neutron fluence values used for the nozzle forgings and the nozzle-to-shell welds are consistent with those listed in WCAP-17441-NP. Furthermore, these neutron fluence values are conservatively based on the nozzle-to-shell weld location closest to the reactor core and were selected based on a 55 EFPY operating term. Therefore, they are

- 9 acceptable. Regarding the other fracture toughness-related parameters, the NRC staff confirmed that the initial RTNDTvalues for the inlet and outlet nozzle forgings, and the Cu content for the inlet nozzle forgings, are consistent with the most limiting values for these forgings, as documented in the Seabrook Updated Final Safety Analysis Report (UFSAR). The NRC staff noted that measured Cu content values for the outlet nozzle forgings are not available. Therefore, the licensee selected a bounding value of 0.40 weight percent. The NRC staff finds this value acceptable because it is more bounding than the 0.35 weight percent value, as recommended by RG 1.99, Rev. 2. The NRC staff noted that the Ni content values are listed in WCAP-17441-NP as being measured values from certified material test reports. Therefore, they are acceptable. For the nozzle-to-shell welds, the NRC staff confirmed that the licensee used the bounding generic Cu and Ni content values recommended by RG 1.99, Rev. 2, as well as the generic initial RTNDTvalue for Linde 0091 submerged arc welds specified in 10 CFR 50.61. Finally, the NRC staff verified that the above ART values were correctly calculated using the procedures in RG 1.99, Rev. 2, Position 1.1. Therefore, the NRC staff finds that the ART values for the inlet and outlet nozzle forgings and associated welds are acceptable. As shown in the table above, the licensee's ART values for the nozzle-to-shell welds are bounded by those for the forgings due to the low generic initial RTNDTvalue invoked for these welds. The licensee also stated that there is no significant stress concentration in these welds, relative to the forged nozzle corner. The NRC staff noted that, although the effects of stress concentration cannot be definitively ruled out for these welds, due to their close proximity to the nozzle discontinuity, the stress concentration effect would be significantly less pronounced than the inside corner of the nozzle forgings. Therefore, the NRC staff finds that the licensee's basis for excluding these welds, in its development of the P-T limit curves for the nozzles, is acceptable. The licensee generated P-T limit curves for the inlet and outlet nozzle forgings using the 55 EFPY ART values listed above, based on a 100 OF per hour cool-down rate and a postulated inside corner flaw of depth 1/4T. The licensee stated that the methods used for determining the applied stress intensity factors due to pressure loading (KIP) and thermal gradients (KIT) are consistent with those published in the Oak Ridge National Laboratory (ORNL) study, ORNLrrM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles - Revision 1," June 2012." The licensee stated that the through-wall stress distributions in the nozzle inside corner region were fitted based on a third-order polynomial of the form given in Equation (1) of ORNLrrM-2010/246. The KIP and KIT values were generated for the postulated 1/4T inside corner flaw, based on the polynomial stress distribution, using the nozzle stress intensity solution provided in Equation (2) of ORNLrrM-201 0/246. The licensee's RAI response included the above stress intensity formula used for analyzing the nozzles' inside corner region. The NRC staff verified that the stress intensity formulation described in the licensee's RAI response has been approved by the NRC for implementation in RPV nozzle KIP and KIT calculations, as documented in the SE for BWR Owners Group Licensing Topical Report SIR 05-044-A, "Pressure-Temperature Limits Report [PTLR] Methodology for Boiling Water 11 ADAMS Accession No. ML12181A162.

- 10 Reactors," Rev. 0, April 2007. 12 The PTLR methodology described in SIR-05-044-A was developed and approved for BWRs seeking to relocate P-T limit curves from the TS to PTLRs. The specific nozzle K,formulation described in the ORNLfTM-2010/246 report is based on a LEFM model that is generally considered to be applicable to postulated corner flaws in rounded corner nozzle forgings, irrespective of plant design. Based on the review of these analyses, the NRC staff finds that the licensee's methods for calculating the K,P and KIT values for the nozzles are acceptable. The licensee noted that an outside surface (3/4T) flaw was not considered in the development of the nozzles' P-T limit curves. The licensee explained that the pressure stress is significantly lower at the outside surface than the inside surface for the nozzles. Additionally, the licensee noted that a stress intensity factor correlation does not exist to consider an outside surface nozzle corner flaw. Based on these considerations, the licensee indicated that nozzle P-T limit curves were not provided for heat-up conditions, as they would be less limiting than the nozzle P-T limit curves for cool-down conditions. The NRC staff determined that the licensee's basis for excluding the 3/4T postulated flaw in the development of the P-T limit curves for the nozzles is acceptable. Specifically, the NRC staff noted that the pressure stress decreases as a function of distance from the inside corner along the through-wall nozzle corner path, as shown in Figure 24 of ORNLfTM-2010/246. Therefore, the K,P value for a 3/4T postulated flaw at the outside corner of the nozzle would be lower than that for the 1/4T flaw postulated for the inside corner region. The ORNL Report ORNLfTM 2010/246 LEFM analyses do not address 3/4T postulated flaws for this reason. It should be noted that, based on the analysis of the 1/4T location, the nozzle P-T limits for heat-up conditions would be less restrictive than those calculated for cool-down conditions, due to the fact that the 1/4T thermal stresses are compressive for heat-up. Therefore, the analysis of the 1/4T location during the cool-down transient generates the most bounding P-T limits for the nozzles. Thus, the NRC staff finds that the licensee's consideration of the nozzle P-T limits for cool-down conditions, based on the analysis of the 1/4T location, is acceptable. The licensee stated that the resulting nozzle P-T limit curves, as provided in the RAI response, are less limiting than the current 20 EFPY (proposed 23.7 EFPY) P-T limits for cool-down conditions developed in WCAP-15745. The licensee indicated that this demonstrates that the nozzle P-T limits are less restrictive, and the limiting RPV beltline shell material is controlling. The licensee also stated that the inlet and outlet nozzles are the most highly stressed ferritic component in the RPV, and therefore, they are the most limiting RPV components with respect to the effects of stress concentration. The NRC staff performed an independent confirmatory calculation to verify that the P-T limits for the inlet and outlet nozzle forgings are bounded by the TS P-T limit curves. The NRC staff's nozzle calculation employed a simplified, but conservative approach by applying a stress concentration factor to the RPV shell membrane stress to account for the elevated stress levels in the nozzle corner region. The NRC staff's confirmatory calculation verified that the inlet and outlet nozzles' P-T limits are less restrictive and therefore bounded by those for the limiting beltline shell material, Lower Shell Plate R-1808-1. Given that the inside corner regions of the inlet and outlet nozzles are the most highly stressed ferritic components in the RPV, the NRC staff determined that the licensee adequately demonstrated that the TS P-T limit curves are 12 ADAMS Accession No. ML070180483.

- 11 controlling for the entire RPV. Therefore, the NRC staff finds that RAI-2 is resolved, as it pertains to factor (1). In its response to RAI-2, factor (2), the licensee stated that the LST requirement of the ASME Code, Section III, NB-2332(b) is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2.5 in. The licensee indicated that the Seabrook RCS does not have ferritic materials in the piping, pumps or valves. The Section III, NB-2332(b) LST requirement only applies to ferritic material in RCPB piping, pumps, and valves with nominal wall thickness greater than 2.5 in. Therefore, the NRC staff finds that the LST requirement of NB-2332(b) is not applicable to the Seabrook P-T limits. The licensee stated that the Seabrook steam generators and pressurizer were designed in accordance with the requirements of the ASME Code, Section III, 1971 Edition through Summer 1973 Addenda. The licensee indicated that the steam generators and pressurizer at Seabrook are all original components. Furthermore, the steam generators and pressurizer do not undergo neutron embrittlement, which is the only aging effect that could impact the P-T limits. The NRC staff determined that, based on the consideration of these components in meeting the original plant design requirements, further consideration of the steam generators and pressurizer for the subject P-T limits LAR is not required. Based on this determination and the LST finding above, the NRC staff finds that RAI-2 is resolved, as it pertains to factor (2). The licensee concluded its response by stating that the RPV nozzles have been addressed for the Seabrook P-T limits and shown to be less limiting than the TS P-T limit curves approved in License Amendment No. 89. Furthermore, the LST requirement of the ASME Code, Section III, NB-2332(b) is not applicable to Seabrook. The other ferritic components of the RCS, namely the steam generators and pressurizer, do not require further evaluation since they were designed to the applicable requirements of the ASME Code, Section III and have not undergone neutron embrittlement. Therefore, the licensee concluded that the P-T limit curve applicability revision proposed in the subject LAR is consistent with the requirements of 10 CFR Part 50, Appendix G. The NRC staff confirmed the licensee's conclusion and found the RAI response acceptable because the licensee has adequately demonstrated that: (1) considering the RPV inlet/outlet nozzles, the TS P-T limit curves are bounding for the entire RPV, and (2) the ferritic RCPB components not part of the RPV were designed to meet the applicable requirements of the ASME Code, Section III, as required by 10 CFR Part 50, Appendix G. Therefore, based on the revised RPV neutron fluence evaluation, the NRC staff finds that the licensee's TS P-T limit curves will continue to meet the requirements of 10 CFR Part 50, Appendix G through 23.7 EFPY. Based on its evaluation, as documented in Section 3.2 of this SE, the NRC staff has determined that the TS P-T limit curves will continue to satisfy the requirements of 10 CFR Part 50,Appendix G through 23.7 EFPY. Therefore, the NRC staff concludes that the proposed 23.7 EFPY applicability term (corresponding to a neutron fluence value of 1.324 x 10" n/cm2 (E > 1.0MeV>> for the P-T limit curves and COPS setpoints is acceptable for incorporation into the Seabrook TS.

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4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Hampshire and Massachusetts State officials were notified of the proposed issuance of the amendment. The State officials provided no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (77 FR 1519). Accordingly, the amendment meets 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 the amendment.

6.0 CONCLUSION

The Commission 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 amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor: Chris Sydnor Date: April 15, 2013

April 15, 2013 Mr. Kevin Walsh Site Vice President clo Michael O'Keefe Seabrook Station NextEra Energy Seabrook, LLC P.O. Box 300 Seabrook, NH 03874

SUBJECT:

SEABROOK STATION, UNIT NO.1-ISSUANCE OF AMENDMENT RE: REVISION TO THE APPLICABILITY OF THE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS AND THE COLD OVERPRESSURE PROTECTION SETPOINTS (TAC NO. ME7645)

Dear Mr. Walsh:

The Commission has issued the enclosed Amendment No. 135 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1 (Seabrook). This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 17,2011, as supplemented December 3,2012, and January 9,2013. The amendment revises the applicability of the figures in the TSs for the reactor coolant system pressure-temperature limits and the cold overpressure protection setpoints. The amendment revises the applicability of the figures from 20 effective full-power years (EFPY) to 23.7 EFPY. A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRA! John G. Lamb, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443

Enclosures:

1. Amendment No. 135 to NPF-86
2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC LPLI-2 RtF RidsAcrsAcnw_MailCTR Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl1-2 Resource RidsRgn1 MailCenter Resource RidsNrrDeEvib Resource RidsNrrPMSeabrook Resource RidsNrrLAABaxter Resource RidsNrrDssSrxb Resource REnnis, NRR/DORL ADAMS Accession No' ML120820510 'via email "via memorandum OFFICE LPL1-2/PM LPL1-2/LA EVIB/BC OGC NLO LPL1-2/BC NAME JLamb ABaxter.... I SRosenberg" MSmith MKhanna DATE . 03/29/2013 414/2013 I 03128 12013 4/912013 4/15/2013 Off., iCial Record Copy}}

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