ML14220A517

County Station, Unit 1, Issuance of Amendment Revising Pressure and Temperature Limits

ML14220A517
Person / Time
Site:	LaSalle
Issue date:	11/25/2014
From:	Purnell B A Plant Licensing Branch III
To:	Pacilio M J Exelon Generation Co, Exelon Nuclear
Blake Purnell, NRR/DORL
References
TAC MF3270
Download: ML14220A517 (18)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 25, 2014 Mr. Michael J. Pacilio Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO) Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNIT 1, ISSUANCE OF AMENDMENT REVISING PRESSURE AND TEMPURATURE LIMITS (TAC NO. MF3270)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 210 to Facility Operating License No. NPF-11 for the LaSalle County Station (LSCS), Unit 1. The amendment is in response to your application dated December 20, 2013, as supplemented by letters dated February 26, September 11 (2 letters), and October 14, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML13358A354, ML14057 A549, ML 14255A348, ML 14258A038, and ML 14288A 151, respectively).

The amendment revises the LSCS, Unit 1, pressure and temperature limit curves, Figures 3.4.11-1 through 3.4.11-3, in technical specification 3.4.11, "RCS [Reactor Coolant System] Pressure and Temperature (PIT) Limits." A copy of the related Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-373

Enclosures:

1. Amendment No. 210 to NPF-11 2. Safety Evaluation cc w/encls: Distribution via ListServ Sincerely, IJ/; iJ-AA-Blake Purnell, Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 21 0 License No. NPF-11 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A The application for amendment to the LaSalle County Station, Unit 1 (the facility)

Operating License No. NPF-11 filed by the Exelon Generation Company, LLC (the licensee), dated December 20, 2013, as supplemented by letters dated February 26, September 11 (2 letters), and October 14, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 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 set forth in 10 CFR Chapter I; 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 page 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-11 is hereby amended to read as follows: (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 210, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Travis L. Tate, Chief Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance:

November 25, 2014 ATTACHMENT TO LICENSE AMENDMENT NO. 210 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Facility Operating License and Appendix "A" Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove License NPF-11 Page 3 TSs 3.4.11-6 3.4.11-7 3.4.11-8 License NPF-11 Page 3 TSs 3.4.11-6 3.4.11-7 3.4.11-8 Am.146 01/12/01 Am. 202 07/21/11 (4) (5) License No, NPF-11 Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 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; and Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of LaSalle County Station, Units 1 and 2 and such Class B and Class C low-level radioactive waste as may be produced by the operation of Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, and Clinton Power Station, Unit 1. 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 thereafter in effect; and is subject to the additional conditions specified or incorporated below: Am. 198 09/16/10 Am. 210 11/ZS/14 Am. 194 08/28/09 Am. 194 08/28/09 Am. 194 08/28/09 Am. 194 08/28/09 Am. 194 08/28/09 ( 1) (2) (3) (4) (5) (6) (7) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3546 megawatts thermal).

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 210, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. DELETED DELETED DELETED DELETED DELETED Amendment No. 210 1400 1300 1200 1100 Ci 1000 c c( w J: 900 D. 0 1-...J w 800 Ill Ill w > a: 700 0 1-0 c( w 600 a: 1-i 500 ::::i w a: ::> 400 Ill Ill w a: D. 300 200 100 0 -----------

-+----

' ' ' ' ' ' ' ' ' -------. ' ' ' ' ' ' ' ' ' I

---

,/----' ' ' ' -----------+-**-BOTTOM HEAD 68'F ' ' ' ' ' ' ' ' .. ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '

J 1/ ---r I I i 1/ i I I / --------' ---------I ' RCS P/T Limits 3.4.11 INITIAL RT NDT VALUES ARE -30'F FOR BEL TUNE, 42'F FOR UPPER VESSEL, AND 47"F FOR BOTTOM HEAD BEL TUNE CURVES ADJUSTED AS SHOWN: EFPY SHIFT ('F) 32 146 HEATUP/COOLDOWN RATE OF COOLANT 20'F/HR 1---------*--I-0 ' ' ' I ' I ' ' ' ' 312 PSIG

-** ----------

FLANGE REGION 1--" .......--72'F ,....... I I 25 50 75 100 125 150 175 200 225 MINIMUM REACTOR VESSEL METAL TEMPERATURE

('F) Figure 3.4.11-1 (Page 1 of 1) Unit 1 --UPPER VESSEL AND BEL TUNE LIMITS -------BOTTOM HEAD CURVE -----------------

-.. P-T Curves for Hydrostatic or Leak Testing up to 32 EFPY i LaSalle 1 and 2 3.4.11-6 Amendment No. 210/188 1400 I 1300 ----'---_____ , -' ' ' ' ' ' ' ' ' ' ' ' ' RCS P/T Limits 3.4 .11 1200 t----1-------i---____ j_ INITIAL RTNDT VALUES ARE 1100 Oi ! 1000 c L5 :I: 900 a.. 0 1-...J w 800 U) U) w t-----t-----t---+---I --:-i I ! o I I I i -=::__[!_-

-3o*F FOR BEL TUNE, 42*F FOR UPPER VESSEL, AND 47"F FOR BOTTOM HEAD BEL TUNE CURVES ADJUSTED AS SHOWN: : I EFPY SHIFT (°F) f---_____ ---1----___ j ----IJ+---+-----I L __ 3_2 __ 1_4_6 __ ...J > a:: 0 1-700 I l I '/ v --() L5 a:: 600 1-i 500 :::i w a:: ::l U) U) w a:: a.. 400 300 200 100 0 0 1/ 1---------,/j -------f-,/ I /I I L_L_ ] I ! --_____ ,_ -------BOTIOM ! HEAD f-:_ _ I --* :f !

• ' ' V REGION t--+--l-It-n*F !_ : It L____.,.______J

! ! !IY I : I 25 50 75 100 125 150 175 200 225 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F) Figure 3.4.11-2 (Page 1 of 1) Unit 1 HEATUP/COOLDOWN RATE OF COOLANT .:: 100.F/HR --UPPER VESSEL AND BEL TUNE LIMITS -------BOTTOM HEAD CURVE P-T Curves for Heatup by Non-Nuclear Means, (Core Not Critical)

Cool down Following a Nuclear Shutdown and Low Power Physics Testing up to 32 EFPY LaSalle 1 and 2 3.4.11-7 Amendment No. 210/188 1400 1300 1200 1100 -.!21 ; 1000 0 ca: w ::t 900 0.. 0 1-...J w 800 If) If) w > a: 700 0 1-0 ca: w 600 a: 1-i 500 :J w a: :::1 400 If) If) w a: 0.. 300 200 100 0 ------------! ' I i I I I ' ! I ! I ! i I i I I I I ! I J i 1-4--------------------------i ----------------------I ------I ! ' i,.--l312 PSIGI I I / ! I Minimum Vessel ! Temperature 72'F I I I I I ! I ' f----! RCS P/T Limits 3. 4.11 INITIAL RT NOT VALUES ARE -30oF FOR BEL TUNE, 42oF FOR UPPER VESSEL, AND 47"F FOR BOTTOM HEAD BEL TUNE CURVE ADJUSTED AS SHOWN: EFPY SHIFT (°F) 32 146 HEATUP/COOLDOWN RATE OF COOLANT _:: 100°F/HR -BEL TUNE AND BEL TUNE LIMITS 0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

("F) Figure 3.4.11-3 (Page 1 of 1) Unit 1 P-T Curves for Operation with a Core Critical other than Low Power Physics Testing up to 32 EFPY LaSalle 1 and 2 3.4.11-8 Amendment No. 210/188 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 210 TO FACILITY OPERATING LICENSE NO. NPF-11

1.0 INTRODUCTION

EXELON GENERATION COMPANY, LLC LASALLE COUNTY STATION, UNIT 1 DOCKET NO. 50-373 By application dated December 20, 2013, as supplemented by letters dated February 26, September 11 (2 letters), and October 14, 2014, (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 13358A354, ML 14057 A549, ML 14255A348, ML 14258A038, and ML 14288A 151, respectively), Exelon Generation Company, LLC (EGC, the licensee) requested changes to the technical specifications (TSs) for the LaSalle County Station (LSCS), Unit 1. The U.S. Nuclear Regulatory Commission (NRC or Commission) staff's proposed no significant hazards consideration determination was published in the Federal Register on August 5, 2014 (79 FR 45490), and referenced the December 20,2013, application and the February 26, 2014, supplement.

Although the supplements dated September 11 and October 14, 2014, superseded some of the information provided in the application, these supplements did not change the technical content of the application since they only clarified what information was considered to be proprietary.

Thus, the supplements dated September 11 and October 14, 2014, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination.

The proposed changes would revise the LSCS, Unit 1, pressure and temperature (P!T) limit curves, Figures 3.4.11-1 through 3.4.11-3, in TS 3.4.11, "RCS [reactor coolant system] Pressure and Temperature (P!T) Limits." The application states that LSCS is participating in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). In February 2010, the 120° capsule was removed from the LSCS, Unit 1, reactor pressure vessel (RPV) and tested in accordance with the BWRVIP protocol of the ISP. Based on the testing, it was determined that the limiting beltline material shift value for Unit 1 had increased which resulted in an increase in the adjusted reference temperature (ART}. As a result, the current LSCS, Unit 1, P!T limit curves are nonconservative.

The proposed TS changes were requested to resolve the nonconservative P/T limit curves. The application states that plant operations in TS 3.4.11 are Enclosure 2 being administrative controlled per NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications that Are Insufficient to Assure Plant Safety." The application also states that the current PIT limit curves for LSCS, Unit 1, have been determined to be adequate through 26.5 effective full-power years (EFPY), and that LSCS, Unit 1, has operated for 21.6 EFPY as of January 1, 2013. The application includes proposed changes to the TS and supporting information for operation to 32 EFPY.

2.0 REGULATORY EVALUATION

The NRC has established requirements in Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50 requires that the PIT limits for the facility's RPV be at least as conservative as those obtained by following the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). The 2010 edition is the most recent version of Appendix G to Section XI of the ASME Code that has been endorsed by the NRC in 10 CFR 50.55a, "Codes and Standards." The 2010 edition of Appendix G to Section XI of the ASME Code incorporates ASME Code Case N-588, "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," and ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of PIT Limit Curves." Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the preservice hydrostatic test pressure.

Appendix G to 10 CFR Part 50, paragraph IV.A, states: "For the reactor vessel beltline materials, including welds, plates and forgings, the values of RT Nor and Charpy upper-shelf energy must account for the effects of neutron radiation, including the results of the surveillance program of appendix H of this part." The effects of neutron radiation are determined, in part, by estimating the neutron fluence on the reactor vessel. Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 10 CFR Part 50 establishes requirements for each facility related to its RPV material surveillance program. Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" (ADAMS Accession No. ML003740284), contains methodologies acceptable to the NRC staff for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity" (ADAMS Accession No. ML031 070438), requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Revision 1, Supplement 1 (ADAMS Accession No. ML031070449), requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (ADAMS Accession No. ML010890301), describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence. Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock," of NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition," (ADAMS Accession No. ML070380185) provides a method acceptable to the NRC staff for determining the PIT limits for ferritic materials in the beltline of the RPV based on the ASME Code, Appendix G, methodology.

3.0 TECHNICAL

EVALUATION The application states that methods described in GE Hitachi Nuclear Energy (GEH) licensing topical report NEDC-33178P-A, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves" (public version under ADAMS Accession No. ML092370487), were used to determine the revised PIT limits for LSCS, Unit 1. A condition of NEDC-33178P-A requires the licensee to calculate the neutron fluence using an NRC-approved methodology.

The calculated neutron fluence is an input to the GEH methodology for determining PIT curves. 3.1 Neutron Fluence Calculation The NRC guidance provided in RG 1.190 indicates that the following attributes comprise an acceptable neutron fluence calculation:

• A fluence calculation performed using an acceptable methodology

• Analytic uncertainty analysis identifying possible sources of uncertainty

• Benchmark comparison to approved results of a test facility

• Plant-specific qualification by comparison to measured fluence values The application states that two different methods for calculating neutron fluence, the Radiation Analysis Modeling Application (RAMA) method and the GEH method, were combined to determine the total fluence value. The neutron fluence for the first 13 operating cycles of LSCS, Unit 1, was calculated in accordance with the RAMA methodology, which is documented in BWRVIP licensing topical report BWRVIP-114P-A, "RAMA Fluence Methodology Theory Manual" (public version available under ADAMS Accession No. ML092650376), and its companion reports.1 The series of four BWRVIP topical reports, and a separate proprietary report documenting the results of a fluence evaluation for a Hope Creek flux wire dosimeter, comprise the RAMA methodology.

A copy of the NRC staff's May 13, 2005, letter approving the RAMA methodology, including the companion reports, is included with BWRVIP-114P-A.

The neutron fluence subsequent to cycle 13 was calculated using GEH Licensing Topical Report NEDC-32983P-A, Revision 2, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," (public version available under ADAMS Accession No. ML072480112).

A copy of the NRC staff's November 17, 2005, letter approving the GEH fluence calculation methodology is included with NEDC-32983P-A.

1 The publicly available versions of the companion reports are BWRVIP-115NP-A, "RAMA Fluence Methodology Benchmark Manual -Evaluation of Regulatory Guide 1.190 Benchmark Problems";

BWRVIP-117NP-A, "RAMA Fluence Methodology Plant Application -Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5"; and BWRVIP-121 NP-A, "RAMA Fluence Methodology Procedures Manual" (ADAMS Accession Nos. ML 100540367, ML 100480185, and ML 100110356, respectively). Section 3.0, "Background," of the application provides the following justification for combining the RAMA and GEH fluence calculation methodologies:

Each NRC approved methodology meets the RG 1.190 requirements and the plant-specific condition of the NRC Safety Evaluation for NEDC-33178P-A.

A comparison of the fluence values for 32 EFPY between the dual calculation (RAMA followed by GEH) and draft calculations of RAMA alone indicates that the dual calculation bounds the single calculation (i.e., results for RAMA alone are less that the dual methodology used in the development of the PIT curves). Therefore, EGC has determined the dual methodology approach utilized to support this LAR [license amendment request] results in more conservative fluence input to the PIT curves. The NRC staff has not previously approved the combination of two methodologies for determining neutron fluence. Furthermore, the guidance provided in RG 1.190 is limited to the use of a single fluence method to determine RPV fluence for the entire irradiation period. The uncertainty analysis and methodology qualification regulatory positions do not provide any guidance for determining an accurate uncertainty estimate or qualifying the fluence estimate used from a combination of methods. As such, the staff concluded that combining fluence values from two separate methods does not adhere to the guidance contained in RG 1.190. Therefore, the staff considered this a deviation from NEDC-33178P-A, and the staff requested a supplement to the application.

In its supplemental letter dated February 26, 2014, the licensee provided a summary of neutron fluence calculations performed using just the RAMA fluence methodology.

These fluence calculations were performed using a single methodology to account for vessel exposure from 0 to 54 EFPY. These calculations confirmed that the neutron fluence estimate used for the proposed PIT limits (i.e., the values obtained from the RAMNGEH combined methodology discussed above) was conservative, since the fluence value calculated using just the RAMA methodology was less than the value obtained using the RAMNGEH combined methodology.

The single-method neutron fluence calculations described in the supplemental letter were performed in a manner consistent with the guidance set forth in RG 1.190. A solution to the Boltzmann transport equation was approximated using the three-dimensional RAMA fluence methodology.

The licensee used the BUGLE-96 wide-group cross section library, which is based on the ENDF/8-VI nuclear data file, in accordance with RG 1.190. Approximations included a P 5/P 7 Legendre expansion for anisotropic scattering, which is in line with the minimum P 3 expansion suggested in RG 1.190. The analytic uncertainty was approximated at 9.2 percent, which is within 20 percent; therefore, the analytic uncertainty is consistent with the guidance in RG 1.190. RAMA is acceptably benchmarked for LSCS, Unit 1. The generic methodology provides general boiling-water reactor (BWR) dosimetry benchmarking.

Specific benchmarking for LSCS, Unit 1, is documented in Electric Power Research Institute, Inc., Technical Report 1022850, "BWRVIP-250NP:

BWRVessel and Internals Project-Testing and Evaluation of the LaSalle Unit 1 120° Surveillance Capsule" (ADAMS Accession No. ML 11326A290).

In addition, the qualification of the methodology included comparisons to the benchmark measurements from the vessel fluence benchmark problems provided in NUREG/CR-6115, "PWR [pressurized-water reactor] and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions." As such, the method is suitably qualified for use in solving BWR fluence problems, including for LSCS, Unit 1. The comparisons discussed above all demonstrated that calculated reaction rates were within 20 percent of measured values, as suggested in RG 1.190; therefore, the NRC staff determined that these benchmarks are acceptable.

As described above, the RAMA fluence calculation methodology, as applied to LSCS, Unit 1, adheres to the guidance in RG 1.190. On this basis, the NRC staff determined that the calculations using just the RAMA fluence methodology are acceptable.

In its supplemental letter dated February 26, 2014, the licensee asserted that the methodology described in NEDC-33178P-A, including the NRC's letter of approval, contains no restriction on the use of two separate fluence methodologies to determine a total fluence. The licensee provided references to documents that it claimed supported this position.

However, the NRC staff did not evaluate this position since the staff had determined, based on the information in the supplemental response, that the neutron fluence value used by the licensee to develop the P/T limit curves based only on the RAMA methodology was conservative.

This safety evaluation shall not be construed as endorsement, agreement with, or approval of, the licensee's position regarding the combined use of fluence methods to determine a total fluence. 3.2 PIT Limit Curves 3.2.1 Licensee's Evaluation The operating limits for PIT are required for three categories of operation:

• hydrostatic pressure tests and leak tests (CurveAorTS Figure 3.4.11-1);

• non-nuclear heatup/cooldown (core not critical) (Curve B or TS Figure 3.4.11-2);

and

• core critical operation (Curve C or TS Figure 3.4.11-3).

The proposed PIT limits are based on application of the GEH methodology in NEDC-33178P-A to LSCS, Unit 1. NEDC-33178P-A provides the NRC-approved generic GEH methodology for generating PIT limits based on the plant-specific ART. NEDC-33178P-A provides generic upper vessel and bottom head PIT limit curves along with beltline curves that are shifted by the specific ART, as well as guidance on the application of Appendix G to the ASME Code (1998 edition and 2000 addenda) and Appendix G to 10 CFR Part 50. For the RPV beltline materials, ART values were calculated for 32 EFPY using plant-specific properties in most cases. The licensee identified weld heat 1 P3571 as the limiting beltline weld material and C6345-1 as the limiting beltline plate for LSCS, Unit 1. The licensee noted that the copper content of the N6 low-pressure coolant injection nozzle weld was conservatively estimated based on BWR fleet and NRC Reactor Vessel Integrity Database data for shielded metal arc weld materials.

Also, the N12 water level instrument (WLI) nozzles were fabricated from Alloy 600, a material that does not require evaluation for fracture toughness; therefore, the WLI nozzles were evaluated using the limiting material properties (chemistry and initial RT Nor) of the adjoining plates. The parameters used to determine the ART values for LSCS, Unit 1, for the limiting materials at the one-quarter of the RPV wall thickness (1/4T) location for 32 EFPY are shown in the PIT limits report (Attachment 2 to the September 11, 2014, letter; ADAMS Accession No. ML 14255A348).

The estimated ART of the limiting weld was significantly above that of the limiting beltline plate. Corresponding parameters at the three-quarter of the RPV wall thickness (3/4T) were not provided.

Instead, the licensee applied the maximum tensile stress for both heatup and cooldown at the 1/4T location.

The licensee stated that this approach is conservative as the 1/4T material toughness is lower than that in the 3/4T locations.

The PIT limits report states: The PIT curves for the non-beltline region were conservatively developed for a [BWR] Product Line 6 (BWR/6) with nominal inside diameter of 251 inches. The analysis is considered appropriate for LSCS Unit 1 because the plant-specific geometric values are bounded by the generic analysis for the large BWR/6. The generic value was adapted to the conditions at LSCS using plant-specific RT NDT values for the [RPV]. The PIT curves specify the maximum coolant heatup and cooldown rates, which is 20 degrees Fahrenheit per hour CF/hr) for Curve A and 100 °F/hr for Curves B and C. However, these curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams.

The PIT limits and corresponding heatup/cooldown rates of either Curve A or B may be applied while achieving or recovering from test conditions.

Curve A applies during pressure testing and when the limits of Curve B cannot be maintained.

The licensee noted that there are six thickness discontinuities in the vessel. Each discontinuity is discussed in the application, as supplemented, and the temperatures used in the development of the PIT curves bound the temperatures associated with the thickness discontinuities.

The PIT limit curves were limited by the beltline curve above 610 pounds per square inch gauge (psig) for Curve A and above 440 psig for Curve B. The licensee provided data from the BWRVIP-135, "BWR Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP) Data Source Book and Plant Evaluations," in compliance with a requirement in the GEH methodology in NEDC-33178P-A.

However, the target plate material of the LSCS, Unit 1, RPV did not match the representative material, so the data from BWRVIP-135 was not used. The BWRVIP-135 source book is used by the industry in compliance with BWRVIP-86, Rev. 1, "BWR Vessel and Internals Project Updated BWR Integrated Surveillance Program Implementation Plan" (ADAMS Accession No. ML090300556).

Information was also included detailing the determination process for evaluating non-beltline, but potentially limiting, components.

3.2.2 NRC Staff's Evaluation To evaluate the licensee's input material property values, the NRC staff first confirmed the licensee's selection of limiting materials.

For LSCS, Unit 1, beltline materials, the staff found that the initial RT NDT. copper, and nickel values were in agreement with information provided by the licensee in a license amendment application dated January 27, 2010 (ADAMS Accession No. ML 100321303), which was approved by the NRC on September 16, 2010 (ADAMS Accession No. ML 101830361

). The licensee reported best estimate chemistry and ISP data from BWRVIP-86, Revision 1, to confirm the collection of credible chemistry and surveillance data. Best estimate chemistries from BWRVIP-86 do not differ from the January 27, 2010, application; therefore, the inclusion of best estimate chemistry does not change the limiting beltline material previously identified by the staff. The licensee only calculated ART values for the RPV 1 /4T location.

The staff determined that using the maximum tensile stress for either heatup or cooldown and applying it at the 1/4T location is equivalent to using the maximum thermal stress intensity factor and the minimum fracture toughness in the heatup and cooldown analysis.

Thus, the proposed PIT limits bound both the heatup and cooldown curves. Based on this, the staff considers the licensee's approach to be acceptable.

The NRC staff independently generated plant-specific PIT limits curves using Appendix G to Section XI of the ASME Code and the same material property and neutron fluence values used by the licensee.

Good agreement was observed between the licensee's proposed composite curves and the curves generated by the staff. In Attachment 3 to the application, the licensee provided the proposed LSCS, Unit 1, TS 3.4.11 PIT limit curves (i.e., the composite PIT limit Curves A, B, and C) which were generated using the GEH methodology.

The licensee's composite curves are consistent with composite curves generated by the NRC staff. The staff applied the GEH methodology by shifting the approved generic GEH bottom curves by the ART for the limiting material identified.

For Curve C, below 20 percent of the hydrostatic test pressure (312.5 psig), the staff found the upper vessel curve generated using the GEH methodology limiting, which is consistent with the composite PIT curve provided by the licensee.

For all other conditions, the requirements in Appendix G to 10 CFR Part 50 for the minimum metal temperature of the closure head flange and vessel flange regions produce limiting "notches", which appear as distinct vertical lines at constant temperature above 312 psig in the licensee's proposed PIT limits. For all LSCS, Unit 1, PIT limit curves, a minimum temperature of 68 oF for the bottom head and 72 oF for the flange region was verified as being ASME Code compliant per the stipulation that these regions must be at least RTNoT + 60 °F. When the pressure is greater than 312 psig, the minimum temperature of 102 oF for Curve A, 132 oF for Curve B, and 172 oF for Curve C are derived (as specified in Appendix G to 10 CFR 50) by adding 90 °F, 120 °F, and 160 °F, respectively, to the RTNoT of 12 oF for the limiting flange material for the three operational conditions.

The NRC staff noted that a beltline N61ow-pressure cooling injection nozzle and beltline N12 WLI nozzles were used in the evaluation of the PIT limit curves. The staff reviewed the dispositioning of these and other relevant nozzles and determined that they were adequately addressed.

The staff also noted that the licensee used Appendix J of the GEH methodology to calculate the PIT limits for the WLI nozzles. The staff has previously determined that the thermal stress value from Appendix J of the GEH methodology is extremely compared to the thermal stress value which would result from the normal and upset transients required to be addressed by the Appendix G to Section XI of the ASME Code-as it is derived from an emergency transient.

Therefore, the resulting applied stress intensity factor determined for the WLI nozzles in the licensee's PIT limit curves are expected to be very conservative.

Based on this, the staff finds the analysis acceptable.

The NRC staff also reviewed the licensee's analysis of non-beltline components and materials contained in the PIT limits report. In many plant designs, the properties of the beltline materials are such that geometric stress concentrations outside the beltline and non-beltline materials may be the limiting factors in portions of the PIT limit curves. Specifically, Section 5.2.3.3.1.6 of the LSCS updated final safety analysis report (ADAMS Accession No. ML 14113A089) states that the preoperational system hydrostatic test for Unit 1 was done at 1563 psig and at a minimum temperature of 118 oF, which is established by the RT Nor of the non-beltline cylinder plate plus 60 °F. The NRC staff review agreed with the licensee's conclusion that there were no non-beltline components and materials that may be the limiting factors in portions of the PIT limit curves; therefore, the licensee's conclusion is acceptable.

3.2 PIT Limit Curves Based on the above evaluation, the NRC staff determined that the licensee's proposed PIT limit curves for LSCS, Unit 1, were generated in accordance with the GEH NEDC-33178P-A methodology, except for the method used to determine the neutron fluence. However, the staff determined that the neutron fluence values used to generate the PIT limit curves were conservative.

Therefore, the staff has determined that the PIT limit curves satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. Therefore, the licensee's proposed PIT limit curves are acceptable for the LSCS, Unit 1, RPV for 32 EFPY. 3.3 Figure 3.4.11-2 Caption The application also proposes to modify the caption for Figure 3.4.11-2 by adding "(Core Not Critical)" in front of "Cooldown Following a Nuclear Shutdown." The NRC staff considers this change to be administrative in nature since it is explanatory and does not change the applicability of Figure 3.4.11-2.

Based on this, the staff determined the change to the Figure 3.4.11-2 caption to be acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendment.

The State official had 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 (79 FR 45490; August 5, 2014). 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 Contributors:

Patrick Purtscher, NRR/EVIB Matthew Hardgrove, NRR/SRXB Date of issuance:

November 25, 2014 Mr. Michael J. Pacilio November 25, 2014 Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO) Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNIT 1, ISSUANCE OF AMENDMENT REVISING PRESSURE AND TEMPURATURE LIMITS (TAC NO. MF3270)

Dear Mr. Pacilio:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 210 to Facility Operating License No. NPF-11 for the LaSalle County Station (LSCS), Unit 1. The amendment is in response to your application dated December 20, 2013, as supplemented by letters dated February 26, September 11 (2 letters), and October 14, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 13358A354, ML 14057A549, ML 14255A348, ML 14258A038, and ML 14288A151, respectively).

The amendment revises the LSCS, Unit 1, pressure and temperature limit curves, Figures 3.4.11-1 through 3.4.11-3, in technical specification 3.4.11, "RCS [Reactor Coolant System] Pressure and Temperature (PIT) Limits." A copy of the related Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-373

Enclosures:

Sincerely, !RAJ Blake Purnell, Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

1. Amendment No. 210 to NPF-11 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsOgcRp Resource RidsRgn3MaiiCenter Resource RidsNrrLASRohrer Resource LPL3-2 R/F RidsNrrDirsltsb Resource RidsNrrDoriDpr Resource Amendment Accession No ML 14220A517 OFFICE LPL3-2/PM LPL3-2/LA SRXB/BC**

NAME BPurnell SRohrer CJackson DATE 10/28/14 10/20/14 4/28/14 OFFICE OGC LPL3-2/BC LPL3-2/PM NAME JWachutka TTate BPurnell DATE 10/30/14 11/25/14 11/25/14 OFFICIAL RECORD COPY RidsNrrPMLaSalle Resource RidsNrrDorllpl3-2 Resource RidsAcrsAcnw_MaiiCTR Resource *via e-mail ** via memo EVIB/BC* STSB/BC SRosenberg REIIiott 9/4/14 9/ 23 /14