ML071690299

License Amendments, Issuance of Amendments Primary Containment Isolation Instrumentation

ML071690299
Person / Time
Site:	LaSalle
Issue date:	06/29/2007
From:	Sands S NRC/NRR/ADRO/DORL/LPLIII-2
To:	Crane C Exelon Generation Co
Sands S,NRR/DORL, 415-3154
Shared Package
ML071690253	List: ML071690299 ML071830450
References
TAC MD0540, TAC MD0541
Download: ML071690299 (18)

Text

June 29, 2007 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS RE: PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION (TAC NOS.

MD0540 AND MD0541)

Dear Mr. Crane:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 182 to Facility Operating License No. NPF-11 and Amendment No. 169 to Facility Operating License No. NPF-18 for the LaSalle County Station, Units 1 and 2, respectively. The amendments are in response to your application dated March 16, 2006, as supplemented by letter dated April 6, 2007.

The amendments revise the allowable values for four reactor core isolation cooling leak detection functions in Technical Specification Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Stephen P. Sands, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosures:

1. Amendment No. 182 to NPF-11

2. Amendment No. 169 to NPF-18

3. Safety Evaluation cc w/encls: See next page

June 29, 2007 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS RE: PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION (TAC NOS.

MD0540 AND MD0541)

Dear Mr. Crane:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 182 to Facility Operating License No. NPF-11 and Amendment No. 169 to Facility Operating License No. NPF-18 for the LaSalle County Station, Units 1 and 2, respectively. The amendments are in response to your application dated March 16, 2006, as supplemented by letter dated April 6, 2007.

The amendments revise the allowable values for four reactor core isolation cooling leak detection functions in Technical Specification Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Stephen P. Sands, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-373 and 50-374

Enclosures:

1. Amendment No.182 to NPF-11

2. Amendment No.169 to NPF-18

3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

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DATE 06/26/07 06/12/07 05/17/07 05/17/07 06/29/07 06/29/07 OFFICIAL RECORD COPY

LaSalle County Station, Units 1 and 2 cc:

Site Vice President - LaSalle County Station U.S. Nuclear Regulatory Commission Exelon Generation Company, LLC 2443 Warrenville Road, Suite 210 2601 North 21st Road Lisle, IL 60532-4352 Marseilles, IL 61341-9757 Illinois Emergency Management Plant Manager - LaSalle County Station Agency Exelon Generation Company, LLC Division of Disaster Assistance &

2601 North 21st Road Preparedness Marseilles, IL 61341-9757 1035 Outer Park Dr Springfield, IL 62704 Manager Regulatory Assurance - LaSalle Exelon Generation Company, LLC Document Control Desk - Licensing 2601 North 21st Road Exelon Generation Company, LLC Marseilles, IL 61341-9757 4300 Winfield Road Warrenville, IL 60555 U.S. Nuclear Regulatory Commission LaSalle Resident Inspectors Office Senior Vice President - Operations Support 2605 North 21st Road Exelon Generation Company, LLC Marseilles, IL 61341-9756 4300 Winfield Road Warrenville, IL 60555 Phillip P. Steptoe, Esquire Sidley and Austin Director - Licensing and Regulatory Affairs One First National Plaza Exelon Generation Company, LLC Chicago, IL 60603 4300 Winfield Road Warrenville, IL 60555 Assistant Attorney General 100 W. Randolph St. Suite 12 Vice President - Regulatory Affairs Chicago, IL 60601 Exelon Generation Company, LLC 4300 Winfield Road Chairman Warrenville, IL 60555 LaSalle County Board 707 Etna Road Associate General Counsel Ottawa, IL 61350 Exelon Generation Company, LLC 4300 Winfield Road Attorney General Warrenville, IL 60555 500 S. Second Street Springfield, IL 62701 Chairman Illinois Commerce Commission 527 E. Capitol Avenue, Leland Building Springfield, IL 62706 Robert Cushing, Chief, Public Utilities Division Illinois Attorney General's Office 100 W. Randolph Street Chicago, IL 60601 Regional Administrator, Region III

LaSalle County Station, Units 1 and 2 cc:

Manager Licensing - Braidwood, Byron, and LaSalle Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Senior Vice President - Midwest Operations Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 182 License No. NPF-11

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated March 16, 2006, as supplemented by letter dated April 6, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:

June 29, 2007 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 182, 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

/RA/

Russell Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: June 29, 2007

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. NPF-18

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by the Exelon Generation Company, LLC (the licensee), dated March 16, 2006, as supplemented by letter dated April 6, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 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. 169, 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

/RA/

Russell Gibbs, Chief Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: June 29, 2007

ATTACHMENT TO LICENSE AMENDMENT NOS. 182 AND 169 FACILITY OPERATING LICENSE NOS. NPF-11 AND NPF-18 DOCKET NOS. 50-373 AND 50-374 Replace the following pages of the Facility Operating Licenses and Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License NPF-11 License NPF-11 Page 3 Page 3 License NPF-18 License NPF-18 Page 3 Page 3 TS TS Page 3.3.6.1-7 Page 3.3.6.1-7

(4) 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 (5) 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.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commissions 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; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3489 megawatts thermal).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 182, 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) Conduct of Work Activities During Fuel Load and Initial Startup The licensee shall review by committee all Unit 1 Preoperational Testing and System Demonstration activities performed concurrently with Unit 1 initial fuel loading or with the Unit 1 Startup Test Program to assure that the activity will not affect the safe performance of the Unit 1 fuel loading or the portion of the Unit 1 Startup Program being performed. The review shall address, as a minimum, system interaction, span of control, staffing, security and health physics, with respect to performance of the activity concurrently with the Unit 1 fuel loading or the portion of the Unit 1 Startup Program being performed. The committee for the review shall be composed of at least three members, knowledgeable in the above areas, and who meet the qualifications for professional-technical personnel specified by Amendment No. 182

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70 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.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commissions 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; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3489 megawatts thermal). Items in Attachment 1 shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 169, 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) Conduct of Work Activities During Fuel Load and Initial Startup The licensee shall review by committee all Unit 2 Preoperational Testing and System Demonstration activities performed concurrently with Unit 2 initial fuel loading or with the Unit 2 Startup Test Program to assure that the activity will not affect the safe performance of the Unit 2 fuel loading or the portion of the Unit 2 Startup Program being performed. The review shall address, as a minimum, system interaction, span of control, staffing, security and health physics, with respect to performance of the activity concurrently with the Unit 2 fuel loading or the portion of the Unit 2 Startup Program being performed. The committee for the review shall be composed of at least three members, knowledgeable in the above areas, and who meet the qualifications for professional-technical personnel specified by section 4.4 of ANSI N18.7-1971. At least one of these three shall be a senior member of the Assistant Superintendent of Operations staff.

Amendment No. 169

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 182 TO FACILITY OPERATING LICENSE NO. NPF-11 AND AMENDMENT NO. 169 TO FACILITY OPERATING LICENSE NO. NPF-18 EXELON GENERATION COMPANY, LLC LASALLE COUNTY STATION, UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC, the Commission) dated March 16, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML060760585), as supplemented by letter dated April 6, 2007 (ADAMS Accession No. ML070990338), Exelon Generation Company, LLC (the licensee), requested changes to the technical specifications (TSs) and facility operating license for LaSalle County Station, Units 1 and 2. The proposed changes would revise the allowable values for four reactor core isolation cooling leak detection functions in TS Table 3.3.6.1-1, Primary Containment Isolation Instrumentation.

The April 6, 2007, supplement, contained clarifying information and did not change the NRC staffs initial proposed finding of no significant hazards consideration.

1.0 BACKGROUND

By letter dated March 16, 2006, supplemented by letter dated April 6, 2007, the licensee proposed a license amendment request (LAR) for LaSalle County Station, Units 1 and 2. The LAR proposes the following TS changes in TS Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, Reactor Core Isolation Cooling (RCIC) leak detection functions:

• Increase the Allowable Value for Function 3.e., RCIC Equipment Room Temperature -

High, from < 291.0EF to < 297.0EF

• Decrease the Allowable Value for Function 3.f., RCIC Equipment Room Differential Temperature - High, from < 189.0EF to < 188.0EF

• Decrease the Allowable Value for Function 3.g., RCIC Steam Tunnel Temperature -

High, from < 277.0EF to < 267.0EF

• Increase the Allowable Value for Function 3.h., RCIC Steam Tunnel Differential Temperature - High, from < 155.0EF to < 163.0EF The licensee stated that while responding to a corrective action plan (CAP) action item, the

existing reactor coolant leak detection calculation, L001324, Revision 5A, and the RCIC leak detection setpoint calculation, NED-I-EIC-0213, Revision 1G, were reviewed in May 2005.

During this review, the licensee found that the current allowable values (AV) in Table 3.3.6.1-1 were derived from calculation L001324, Revision 0, rather than Revision 5A, and are therefore, no longer supported by the current calculation NED-I-EIC-0213, Revision 1G. The licensees proposed TS changes are based on revised analytical limits for a leak rate equivalent to 25 gallons per minute determined by L001324, Revision 5A.

2.0 REGULATORY EVALUATION

The NRC staff used the following regulatory bases and guidance documents in its evaluation of the LAR:

2.1 10 CFR Part 50.36, Technical Specifications.

Section 50.36(a) states, Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specification in accordance with the requirements of this section.

Section 50.36(c)(1)(i)(A) states, Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of physical barriers that guard against the uncontrolled release of radioactivity.

Section 50.36(c)(1)(ii)(A) states, Limiting safety system settings [LSSS] for nuclear reactor are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

Section 50.36(c)(3) states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

2.2 NRC Regulatory Issue Summary (RIS) 2006-17, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specifications, Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels, discusses the NRCs requirements on limiting safety system settings (LSSSs) assessed during periodic testing and calibration of instrumentation. This RIS discusses issues that could occur during testing of LSSSs and which, therefore, may have adverse effect on equipment operability and an acceptable approach for addressing these issues via licensing actions that require prior NRC staff approval. This document is available on the NRC public website at ADAMS Accession No. ML051810077.

2.3 Letter from Patrick L. Hiland, NRC, to the Nuclear Energy Institute Setpoint Methods Task Force, Technical Specification for Addressing Issues Related to Setpoint Allowable Values, dated September 7, 2005 (the Hiland letter), ADAMS Accession

No. ML052500004. This letter addresses the TS Notes required for the TS setpoint changes involving safety limit-related instrumentations and corrective actions required when the as-found setpoints are beyond the acceptable as-found setpoints.

2.4 Regulatory Guide (RG) 1.105, Revision 3, Setpoints for Safety-Related Instrumentation, describes a method acceptable to the NRC staff for complying with the NRCs regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits.

2.5 The International Society for Measurements and Control Standard, ISA-S67.04, Part I, Setpoints for Nuclear Safety-Related Instrumentation,, defines a framework for ensuring that setpoints for nuclear safety limit-related instrumentation are established and maintained within specified limits. RG 1.105 endorses ISA-S67.04, Part I, subject to NRC staff clarifications.

3.0 TECHNICAL EVALUATION

The RCIC system provides relatively cool water to the reactor vessel either from the condensate storage tank or from the suppression pool and uses steam from the reactor to drive the RCIC turbine driven pump. The RCIC system is automatically initiated to ensure adequate core cooling in the event the reactor is isolated from the main condenser during power operation with a loss of main feedwater flow. Reactor vessel water level is maintained or supplemented by the RCIC system during the following conditions:

a. reactor vessel isolated and maintained in the hot standby condition;

b. reactor vessel isolated coincident with loss of normal coolant flow from the reactor feedwater system; or

c. plant shutdown where a loss of normal feedwater system is experienced before the reactor is depressurized to where the reactor shutdown cooling mode of the residual heat removal system can be placed in operation.

The licensee stated that the RCIC leak detection isolation functions are not credited in any Updated Final Safety Analysis Report (UFSAR) transient or accident analysis, because the bounding analyses are performed for large breaks such as reactor recirculation line or main steam line breaks. The RCIC system is a safe shutdown system rather than an emergency core cooling system. Therefore, no safety limits are associated with the RCIC system as defined in 10 CFR 50.36(c)(1)(i)(A).

The Hiland letter states that TS Notes should be added for setpoint verification surveillances for instrument functions on which a safety limit has been placed. As stated previously, there are no safety limits associated with the RCIC system, therefore, the NRC staff concludes that there is no need to put the referenced TS Notes in the TS for the proposed AV changes in TS Table 3.3.6.1-1.

The Hiland letter also states that operability and expected performance of instruments will be confirmed during performance of surveillance tests. The surveillance should be conducted as outlined below:

1. If the as-found trip set point (TSP) is found to be non-conservative with respect to the AV specified in TSs, the channel is declared inoperable and the associated TS action statement must be followed.

2. If the as-found TSP is found to be conservative with respect to the AV, and outside the as-found predefined acceptance criteria band, but the licensee is able to determine that the instrument channel is functioning as required and the licensee can reset the channel to within the setting tolerance of the limiting TSP, or a value more conservative than the limiting TSP, then the licensee may consider the channel operable. If the licensee cannot determine that the instrument channel is functioning as required, the channel is declared inoperable and the associated TS actions must be followed.

3. If the as-found TSP is outside the as-found predefined acceptance criteria band, the condition must be entered into the licensees corrective action program for further evaluation.

In response to the NRC staff request for additional information, by letter dated April 6, 2007, the licensee provided copies of the following surveillance procedures:

1. LIS-RI-1(2)03A, Unit 1(2) RCIC Equipment Room/Steam Line Tunnel High Ambient and Differential Temperature Outboard Isolation (DIV 1) Calibration.

2. LIS-RI-1(2)03B, Unit 1(2) RCIC Equipment Room/Steam Line Tunnel High Ambient and Differential Temperature Outboard Isolation (DIV 2) Calibration.

The licensee described the measures taken to ensure instrument channels are operating properly in the response to RAI Question No. 2. These procedures contain a statement to verify that the as-found data is within the required calibration limits. If the as-found data is not within the required calibration limits, the applicable procedure directs the instrument technician to identify the as-found data and contact the Instrument Maintenance Supervisor for further instructions and, if appropriate, calibrate the instrument and record the as-left data.

The licensee, also, provided procedure ER-AA-520, Instrument Performance Trending which details the plant procedures followed when the instruments are found outside the calibration limits. An Issue Report (IR) is initiated when any instrument is found out of calibration limits during periodic surveillances. An IR is the entry point into the plant CAP. If an instrument is found to be outside the as-found calibration limits (acceptable as-found) and outside the AV, the calibration information is documented in the IR. The CAP is also the tracking process to ensure the condition is evaluated, corrected, and the appropriate resolution documented.

ER-AA-520 also documents the requirements for reporting instruments that are found outside the as-found calibration limits but within AV. The calibration information is documented and entered into the CAP as a Condition Report for periodic trending evaluation. All instruments are

required to be left within the as-left calibration limits by ER-AA-520, LIS-RI-1(2)03A and LIS-RI-1(2)03B.

The NRC staff reviewed the information provided in the licensees RAI response. The NRC staff concludes that the licensee has adequate plant procedures to assure that surveillances are conducted in accordance with the three criteria detailed in the Hiland letter. Therefore, the NRC staff finds that the operability and expected performance of instruments will be confirmed during performance of surveillance tests.

In response to staff request for additional information by letter dated April 6, 2007, the licensee provided:

1. Engineering Calculation NED-I-EIC-0213, RCIC Equipment Area/Pipe Tunnel High Ambient and Differential Temperature Outboard and Inboard Isolation Error Analysis, Revision 00

2. Engineering Calculation NED-I-EIC-0213, Derived Technical Specification Allowable Values for Instrument Loop Channels that detect RCIC Equipment Area/Pipe Tunnel High Ambient and Differential Temperature based on Corrected Values for Analytical Limits, Revision 1G

3. Engineering Standard NES-EIC-20.04, Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy. Revision 4.

By letter dated March 16, 2006, the licensee, also, provided NES-EIC-20.04, Equations for Instrument Channel Uncertainties, Setpoints and Allowable Values, Revision 4.

From the above calculations, the licensee arrived at the following setpoint parameters:

TABLE 1: SETPOINT PARAMETERS Function No. in Table 3.3.6.1-1 3e 3f 3g 3h Analytical Limit (AL) 299.6EF 195.6EF 270.5EF 170.5EF Total Error (TE) 11.244EF 8.246EF 11.244EF 8.246EF Channel Drift (DTI) 8.739EF 1.298EF 8.739EF 1.298EF Acceptable As-Found Tolerance (ET) +7EF +1.5EF +7EF +1.5EF Acceptable As-Left Tolerance (ST) +3EF +1.5EF +3EF +1.5EF Trip Setpoint (TS) 192.0EF 108.5EF 192.0EF 108.5EF Limiting Setpoint (SP) 288.356EF 187.354EF 259.256EF 162.254EF Allowable Value (AV) 297.0EF 188.0EF 267.0EF 163.0EF Margin (SP - TS) 96.356EF 78.854EF 53.256EF 53.754EF Margin (AV - TS - ET) 98.0EF 78.0EF 98.0EF 53E

As indicated in Calculation Standard NES-EIC-20.04, Revision 4, in calculating TE the licensee included:

1. Instrumentation calibration uncertainties: Including calibration standards, calibration measurement and test equipment error, and setting tolerances

2. Calibration methods

3. Instrument uncertainties during normal operation: Including reference accuracy, power supply voltage and frequency changes, ambient temperature changes, humidity changes, pressure changes, in service vibration allowances, radiation exposure, analog to digital and digital to analog conversion

4. Instrument drift

5. Uncertainties caused by design basis events

6. Process dependent effects

7. Calculation effects

8. Dynamic effects

9. Installation biases The licensee provided calculations of the TE, DTI, ST, and ET in calculations NED-I-EIC-0213, Revision 00, and NED-I-EIC-0213, Revision 1G. The licensees calculations demonstrated 95 percent/95 percent confidence level in uncertainty calculations. The NRC staff has reviewed the parameters used in calculating TE and find them acceptable, specifically considering the wide margins the licensee used, between 53 EF and 96 EF, for calculating TS from AL, compared to TE of 11.2 EF and 8.2 EF. Similarly, for calculating the AV the licensee used margin between 53E F and 98EF, while the DTIs are 8.7 EF and 1.3 EF. The staff, specifically, checked the derivation of the acceptable as-left and acceptable as-found values and find acceptable as-left values of 3 EF and 1.5 EF and acceptable as-found values of 7 EF and 1.5 EF, respectively, as reasonably low and, therefore, acceptable. Specifically, the channel as-found values are below the DTI values.

Because the licensee used wide margins in calculating the trip setpoint and has used reasonable parameters in calculating the acceptable as-left and acceptable as-found values, the NRC staff finds the calculations acceptable. In addition, the licensee described the plant procedures to set the channels within acceptable as-left values or declare the channel INOPERABLE, where required, and enter into adequate corrective actions. Based on the above considerations, the NRC staff finds that the revised TS values will ensure the RCIC system will perform its intended function.

The NRC staff finds that the proposed TS changes are not related to any plant safety limits and accordingly, there is no need to add the two TS Notes recommended in the NRC letter dated September 7, 2005, for safety limit-related setpoint TS changes. Furthermore, the NRC staff

finds the acceptable as-left and acceptable as-found values the licensee selected are reasonable compared to the wide margin the licensee used in calculating the trip setpoints.

The NRC staff, also, finds the licensee has satisfactory plant procedures to take adequate corrective actions when the setpoints are found beyond the acceptable as-found values during the surveillance tests. In considerations of the above, the NRC staff concludes that the proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change surveillance requirements with respect to the installation or use of the facilities components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (71 FR 46929; August 15, 2006). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: S. Mazumdar Date: June 29, 2007