ML081350179

Issuance of Amendments Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specifications

ML081350179
Person / Time
Site:	Indian Point   (DPR-026, DPR-064)
Issue date:	06/17/2008
From:	Boska J NRC/NRR/ADRO/DORL/LPLI-1
To:	Entergy Nuclear Operations
Boska J, NRR, 301-415-2901
References
TAC MD7564, TAC MD7565, FOIA/PA-2016-0148
Download: ML081350179 (21)

Text

June 17, 2008 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 - ISSUANCE OF AMENDMENTS RE: DELETION OF E BAR DEFINITION AND REVISION TO REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY TECHNICAL SPECIFICATIONS (TAC NOS. MD7564 AND MD7565)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 254 to Facility Operating License No.

DPR-26 for the Indian Point Nuclear Generating Unit No. 2 and Amendment No. 237 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated December 20, 2007.

The amendments revise the TSs associated with Reactor Coolant System (RCS) Specific Activity and the deletion of the TS definition of E Bar (average disintegration energy) consistent with Revision 0 to TS Task Force (TSTF) Standard Technical Specification Change Document TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on March 15, 2007 (72 FR 12217).

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely,

/RA/

John P. Boska, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286

Enclosures:

1. Amendment No. 254 to DPR-26

2. Amendment No. 237 to DPR-64

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

June 17, 2008 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 - ISSUANCE OF AMENDMENTS RE: DELETION OF E BAR DEFINITION AND REVISION TO REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY TECHNICAL SPECIFICATIONS (TAC NOS. MD7564 AND MD7565)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 254 to Facility Operating License No.

DPR-26 for the Indian Point Nuclear Generating Unit No. 2 and Amendment No. 237 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated December 20, 2007.

The amendments revise the TSs associated with Reactor Coolant System (RCS) Specific Activity and the deletion of the TS definition of E Bar (average disintegration energy) consistent with Revision 0 to TS Task Force (TSTF) Standard Technical Specification Change Document TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec. A notice of availability for this TS improvement using the consolidated line item improvement process was published in the Federal Register on March 15, 2007 (72 FR 12217).

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely,

/RA/

John P. Boska, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286

Enclosures:

1. Amendment No. 254 to DPR-26

2. Amendment No. 237 to DPR-64

3. Safety Evaluation cc w/encls: See next page Package No.: ML081350219 Amendment No.: ML081350179 Tech Spec No.: ML

• See Safety Evaluation dated April 1, 2008 OFFICE LPL1-1/PM LPL1-1/LA ITSB/BC OGC LPL1-1/BC NAME JBoska SLittle RElliot*

STurk MKowal DATE 5/27/08 5/28/08 4/1/2008 5/28/08 6/16/08 Official Record Copy

DATED: June 17, 2008 AMENDMENT NO. 254 TO FACILITY OPERATING LICENSE NO. DPR-26 INDIAN POINT UNIT 2 AND AMENDMENT NO. 237 TO FACILITY OPERATING LICENSE NO. DPR-64 INDIAN POINT UNIT 3 PUBLIC LPL1-1 R/F RidNrrDorl RidsNrrDorlLpl1-1 RidsNrrDorlDpr RidsNrrDirsItsb RidsNrrPMJBoska RidsNrrLASLittle (paper copy)

RidsRgn1MailCenter GHill (4) (paper copies)

RidsOGCRp RidsAcrsAcnw&mMailCenter cc: Plant Mailing list

Indian Point Generating Unit Nos. 2 and 3 cc:

Senior Vice President Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Vice President Oversight Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Senior Manager, Nuclear Safety & Licensing Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Senior Vice President and COO Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Assistant General Counsel Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Manager, Licensing Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Mr. Paul Tonko President and CEO New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. John P. Spath New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector=s Office Indian Point 2 U. S. Nuclear Regulatory Commission P.O. Box 59 Buchanan, NY 10511 Senior Resident Inspector=s Office Indian Point 3 U. S. Nuclear Regulatory Commission P.O. Box 59 Buchanan, NY 10511 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Raymond L. Albanese Four County Coordinator 200 Bradhurst Avenue Unit 4 Westchester County Hawthorne, NY 10532 Mayor, Village of Buchanan 236 Tate Avenue Buchanan, NY 10511 Mr. Jim Riccio Greenpeace 702 H Street, NW Suite 300 Washington, DC 20001

Indian Point Generating Unit Nos. 2 and 3 cc:

Mr. Phillip Musegaas Riverkeeper, Inc.

828 South Broadway Tarrytown, NY 10591 Mr. Mark Jacobs IPSEC 46 Highland Drive Garrison, NY 10524 Mr. Sherwood Martinelli FUSE USA via email

ENTERGY NUCLEAR INDIAN POINT 2, LLC ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 254 License No. DPR-26

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated December 20, 2007 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-26 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 254 are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Mark G. Kowal, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: June 17, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 254 FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following pages of the 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 Pages Insert Pages 1.1-2 1.1-2 3.4.16-1 3.4.16-1 3.4.16-2 3.4.16-2

ENTERGY NUCLEAR INDIAN POINT 3, LLC ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 237 License No. DPR-64

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated December 20, 2007 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-64 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 237, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Mark G. Kowal, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: June 17, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 237 FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following pages of the 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 Pages Insert Pages 1.1-3 1.1-3 3.4.16-1 3.4.16-1 3.4.16-2 3.4.16-2 3.4.16-3 3.4.16-4

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 254 TO FACILITY OPERATING LICENSE NO. DPR-26 AND AMENDMENT NO. 237 TO FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

1.0 INTRODUCTION

By letter dated December 20, 2007, Agencywide Documents Access and Management System (ADAMS) Accession No. ML073650053, Entergy Nuclear Operations, Inc. (the licensee) submitted a request for changes to the Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3) Technical Specifications (TSs). The requested changes are the adoption of TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to Reactor Coolant System (RCS) Specific Activity TS for pressurized water reactor (PWR) Standard Technical Specifications (STS).

By letter dated September 13, 2005, the Technical Specification Task Force (TSTF) submitted TSTF-490 for Nuclear Regulatory Commission (NRC) staff review. This TSTF involves changes to NUREG-1430, NUREG-1431, and NUREG-1432 STS Section 3.4.16 RCS gross specific activity limits with the addition of a new limit for noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-133 (DEX) definition that replaces the current E Bar average disintegration energy definition. In addition, the current dose equivalent I-131 (DEI) definition would be revised to allow the use of additional thyroid dose conversion factors (DCFs). The model safety evaluation prepared by the Nuclear Regulatory Commission (NRC) was initially published in the Federal Register on November 20, 2006 (71 FR 67170) under the consolidated line item improvement process (CLIIP), and revised on March 15, 2007 (72 FR 12217).

Entergy is proposing a site-specific wording preference for the definition of Dose Equivalent I-131. The definition for Dose Equivalent Xe-133, as stated in the TSTF and as proposed for adoption at IP2 and IP3, includes a statement regarding the use of minimum detectable activity (MDA). Entergy proposes to also include this statement in the definition of Dose Equivalent I-131. There is no technical reason for the two terms to be treated differently in this regard, thus the NRC staff finds this change acceptable.

2.0 REGULATORY EVALUATION

The NRC staff evaluated the impact of the proposed changes as they relate to the radiological consequences of affected design-basis accidents (DBAs) that use the RCS inventory as the source term. The source term assumed in radiological analyses should be based on the activity

associated with the projected fuel damage or the maximum RCS TS values, whichever maximizes the radiological consequences. The limits on RCS specific activity ensure that the offsite doses are appropriately limited for accidents that are based on releases from the RCS with no significant amount of fuel damage.

The Steam Generator Tube Rupture (SGTR) accident and the Main Steam Line Break (MSLB) accident typically do not result in fuel damage and therefore the radiological consequence analyses are based on the release of primary coolant activity at maximum TS limits. For accidents that result in fuel damage, the additional dose contribution from the initial activity in the RCS is not normally evaluated and is considered to be insignificant in relation to the dose resulting from the release of fission products from the damaged fuel.

For licensees that incorporate the source term as defined in Technical Information Document (TID) 14844, U.S. Atomic Energy Commission (AEC), 1962, Calculation of Distance Factors for Power and Test Reactors Sites, in their dose consequence analyses, the NRC staff uses the regulatory guidance provided in NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.1.5, Steam System Piping Failures Inside and Outside of Containment (PWR), Appendix A, Radiological Consequences of Main Steam Line Failures Outside Containment, Revision 2, for the evaluation of MSLB accident analyses and NUREG-0800, SRP Section 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR), Revision 2, for evaluating SGTR accidents analyses.

In addition, the NRC staff uses the guidance from Regulatory Guide (RG) 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light Water Nuclear Power Reactors, May 2003, for those licensees that chose to use its guidance for dose consequence analyses using the TID 14844 source term.

For licensees using the alternative source term (AST) in their dose consequence analyses, such as IP2 and IP3, the NRC staff uses the regulatory guidance provided in NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, July 2000, and the methodology and assumptions stated in RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

The applicable dose criterion for the evaluation of DBAs depends on the source term incorporated in the dose consequence analyses. For licensees using the TID 14844 source term, the maximum dose criteria to the whole body and the thyroid that an individual at the exclusion area boundary (EAB) can receive for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, and at the low population zone (LPZ) outer boundary for the duration of the radiological release, are specified in Title 10 of the Code of Federal Regulations (10 CFR) Section 100.11. These criteria are 25 roentgen equivalent man (rem) total whole body dose and 300 rem thyroid dose from iodine exposure. The accident dose criteria in 10 CFR 100.11 are supplemented by accident specific dose acceptance criteria in SRP 15.1.5, Appendix A, SRP 15.6.3 or Table 4 of RG 1.195.

For control room dose consequence analyses that use the TID 14844 source term, the regulatory framework on which the NRC staff bases its acceptance is General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50, Control Room. GDC 19 requires that adequate radiation protection be provided to permit access and occupancy of the control room under

accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. NUREG-0800, SRP Section 6.4, Control Room Habitability System, Revision 2, July 1981, provides guidelines defining the dose equivalency of 5 rem whole body as 30 rem for both the thyroid and skin dose.

For licensees adopting the guidance from RG 1.196, Control Room Habitability at Light Water Nuclear Power Reactors, May 2003, an annual organ dose limit of 50 rem can be used for both the thyroid and skin dose equivalent of 5 rem whole body. This is defined in Section C.4.5 of RG 1.195, which states that in lieu of the dose equivalency guidelines from Section 6.4 of NUREG-0800, the 10 CFR 20.1201 limits may be used.

Licensees using the AST, such as IP2 and IP3, are evaluated against the dose criteria specified in 10 CFR 50.67(b)(2). The off-site dose criteria are 25 rem total effective dose equivalent (TEDE) at the EAB for any 2-hour period following the onset of the postulated fission product release and 25 rem TEDE at the outer boundary of the LPZ for the duration of the postulated fission product release. In addition, 10 CFR 50.67(b)(2)(iii) requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

3.0 TECHNICAL EVALUATION

3. 1 Technical evaluation of TSTF-490 TS changes 3.1.1 Revision to the Definition of Dose Equivalent Iodine (DEI)

The Dose Conversion Factors (DCFs) for use in the determination of DEI at IP2 and IP3 shall be the Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) DCFs from Table 2.1 of EPA Federal Guidance Report No. 11, as these are used by the licensee in the dose consequence analyses.

3.1.2 Deletion of the Definition of E Bar and the Addition of a New Definition for DE Xe-133 The new definition for DEX is similar to the definition for DEI. The determination of DEX will be performed in a similar manner to that currently used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 which are significant in terms of contribution to whole body dose. Some noble gas isotopes are not included due to low concentration, short half life, or small dose conversion factor. The calculation of DEX at IP2 and IP3 will use the effective dose conversion factors from Table III.1 of Environmental Protection Agency (EPA) FGR No. 12. Using this approach, the limit on the amount of noble gas activity in the primary coolant would not fluctuate with variations in the calculated values of E Bar. If a specified noble gas nuclide is not detected, the new definition states that it should be assumed the nuclide is present at the minimum detectable activity. This will result in a conservative calculation of DEX.

When E Bar is determined using a design basis approach in which it is assumed that 1.0% of the power is being generated by fuel rods having cladding defects and it is also assumed that there is no removal of fission gases from the letdown flow, the value of E Bar is dominated by

Xe-133. The other nuclides have relatively small contributions. However, during normal plant operation there are typically only a small amount of fuel clad defects and the radioactive nuclide inventory can become dominated by tritium and corrosion and/or activation products, resulting in the determination of a value of E Bar that is very different than would be calculated using the design basis approach. Because of this difference, the accident dose analyses become disconnected from plant operation and the TS limiting condition for operation (LCO) becomes essentially meaningless. It also results in a TS limit that can vary during operation as different values for E Bar are determined.

This change will implement an LCO that is consistent with the whole body radiological consequence analyses which are sensitive to the noble gas activity in the primary coolant but not to other non-gaseous activity currently captured in the E Bar definition. LCO 3.4.16 specifies the limit for primary coolant gross specific activity as 100/E Bar Ci/gm. The current E Bar definition includes radioisotopes that decay by the emission of both gamma and beta radiation. The current Condition B of LCO 3.4.16 would rarely, if ever, be entered for exceeding 100/E Bar since the calculated value is very high (the denominator is very low) if beta emitters such as tritium (H-3) are included in the determination, as required by the E Bar definition. TS Section 1.1 definition for E - AVERAGE DISINTEGRATION ENERGY (E Bar) is deleted and replaced with a new definition for DEX which states:

DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr 87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe 138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE 133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

The change incorporating the newly defined quantity DEX is acceptable from a radiological dose perspective since it will result in an LCO that more closely relates the non-iodine RCS activity limits to the dose consequence analyses which form their bases.

3.1.3 LCO 3.4.16, RCS Specific Activity LCO 3.4.16 is modified to specify that iodine specific activity in terms of DEI and noble gas specific activity in terms of DEX shall be within limits. Currently the limiting indicators are not explicitly identified in the LCO, but are instead defined in current Condition C and Surveillance Requirement (SR) 3.4.16.1 for gross non-iodine specific activity and in current Condition A and SR 3.4.16.2 for iodine specific activity.

The change states RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

3.1.4 TS 3.4.16 Applicability TS 3.4.16 Applicability is modified to include all of MODE 3 and MODE 4. It is necessary for the LCO to apply during MODES 1 through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these MODES. In MODE 5 with the RCS loops filled, the

steam generators are specified as a backup means of decay heat removal via natural circulation. In this mode, however, due to the reduced temperature of the RCS, the probability of a DBA involving the release of significant quantities of RCS inventory is greatly reduced.

Therefore, monitoring of RCS specific activity is not required. In MODE 5 with the RCS loops not filled and in MODE 6 the steam generators are not used for decay heat removal, the RCS and steam generators are depressurized and primary to secondary leakage is minimal.

Therefore, the monitoring of RCS specific activity is not required. The change to modify the TS 3.4.16 Applicability to include all of MODE 3 and MODE 4 is necessary to limit the potential radiological consequences of an SGTR or MSLB that may occur during these MODES and is therefore acceptable from a radiological dose perspective.

3.1.5 TS 3.4.16 Condition A TS 3.4.16 Condition A is revised by replacing the DEI site specific limit > 1.0 Ci/gm with the words not within limit to be consistent with the revised TS 3.4.16 LCO format. The site specific DEI limit of 1.0 Ci/gm is contained in SR 3.4.16.2. This proposed format change will not alter current TS requirements and is acceptable from a radiological dose perspective. TS 3.4.16 Required Action A.1 is revised for IP3 TS only to remove the reference to Figure 3.4.16-1 Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit versus Percent of RATED THERMAL POWER and insert a limit of less than or equal to the site specific DEI spiking limit.

The curve contained in Figure 3.4.16-1 was provided by the AEC in a June 12, 1974, letter from the AEC on the subject, Proposed Standard Technical Specifications for Primary Coolant Activity. Radiological dose consequence analyses for SGTR and MSLB accidents that take into account the pre-accident iodine spike do not consider the elevated RCS iodine specific activities permitted by Figure 3.4.16-1 for operation at power levels below 80% RTP. Instead, the pre-accident iodine spike analyses assume a DEI concentration 60 times higher than the corresponding long term equilibrium value, which corresponds to the specific activity limit associated with 100% RTP operation. It is acceptable that TS 3.4.16 Required Action A.1 should be based on the short term site specific DEI spiking limit to be consistent with the assumptions contained in the radiological consequence analyses.

3.1.6 TS 3.4.16 Condition B Revision to include Action for DEX Limit TS 3.4.16 Condition B is replaced with a new Condition B for DEX not within limits. This change is made to be consistent with the change to the TS 3.4.16 LCO which requires the DEX specific activity to be within limits as discussed above in Section 3.1.3. The DEX limit is site specific and the numerical value in units of Ci/gm is contained in revised SR 3.4.16.1. The site specific limit of DEX in Ci/gm is established based on the maximum accident analysis RCS activity corresponding to 1% fuel clad defects with sufficient margin to accommodate the exclusion of those isotopes based on low concentration, short half life, or small dose conversion factors. The primary purpose of the TS 3.4.16 LCO on RCS specific activity and its associated Conditions is to support the dose analyses for DBAs. The whole body dose is primarily dependent on the noble gas activity, not the non-gaseous activity currently captured in the E Bar definition.

The Completion Time for revised TS 3.4.16 Required Action B.1 will require the restoration of DEX to within the specified limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This is consistent with the Completion Time for current Required Action A.2 for DEI. The radiological consequences for the SGTR and the MSLB accidents demonstrate that the calculated thyroid doses are generally a greater percentage of the applicable acceptance criteria than the calculated whole body doses. It then

follows that the Completion Time for noble gas activity being out of specification in the revised Required Action B.1 should be at least as great as the Completion Time for iodine specific activity being out of specification in current Required Action A.2. Therefore, the Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for revised Required Action B.1 is acceptable from a radiological dose perspective.

A Note is also added to the revised Required Action B.1 that states LCO 3.0.4.c is applicable.

This Note would allow entry into a Mode or other specified condition in the LCO Applicability when LCO 3.4.16 is not being met and is the same Note that is currently stated for Required Actions A.1 and A.2. The proposed Note would allow entry into the applicable Modes from MODE 4 to MODE 1 (power operation) while the DEX limit is exceeded and the DEX is being restored to within its limit. This Mode change is acceptable due to the significant conservatism incorporated into the DEX specific activity limit, the low probability of an event occurring which is limiting due to exceeding the DEX specific activity limit, and the ability to restore transient specific excursions while the plant remains at, or proceeds to power operation.

3.1.7 TS 3.4.16 Condition C TS 3.4.16 Condition C is revised to include Condition B (DEX not within limit) if the Required Action and associated Completion Time of Condition B is not met. This is consistent with the changes made to Condition B which now provide the same Completion Time for both components of RCS specific activity as discussed in the revision to Condition B. For IP3 TS only, the revision to Condition C also replaces the limit on DEI from the deleted Figure 3.4.16-1, with a site-specific value of > 60 Ci/gm. This change makes Condition C consistent with the changes made to TS 3.4.16 Required Action A.1.

The change to TS 3.4.16 Required Action C.1 requires the plant to be in MODE 3 with Tavg <

500 oF within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and adds a new Required Action C.2 which requires the plant to be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Although the TSTF-490 changes do not require Tavg < 500 oF in Required Action C.1, the licensee has decided to maintain this requirement in the adoption of TSTF-490. This variation is more conservative than the change allowed for by the TSTF-490, and thus, the NRC staff finds it acceptable.

The remainder of the changes are consistent with the changes made to the TS 3.4.16 Applicability. The revised LCO is applicable throughout all of MODES 1 through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these MODES.

In MODE 5 with the RCS loops filled, the steam generators are specified as a backup means of decay heat removal via natural circulation. In this mode, however, due to the reduced temperature of the RCS, the probability of a DBA involving the release of significant quantities of RCS inventory is greatly reduced. Therefore, monitoring of RCS specific activity is not required.

In MODE 5 with the RCS loops not filled and MODE 6, the steam generators are not used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.

A new TS 3.4.16 Required Action C.2 Completion Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is added for the plant to reach MODE 5. This Completion Time is reasonable, based on operating experience, to reach MODE 5 from full power conditions in an orderly manner and without challenging plant systems and the value of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is consistent with other TS which have a Completion Time to reach MODE 5.

3.1.8 SR 3.4.16.1 DEX Surveillance The change replaces the current SR 3.4.16.1 surveillance for RCS gross specific activity with a surveillance to verify that the site specific reactor coolant DEX specific activity is 632 Ci/gm for IP2 and 652 Ci/gm for IP3. This change provides a surveillance for the new LCO limit added to TS 3.4.16 for DEX. The revised SR 3.4.16.1 surveillance requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days, which is the same frequency required under the current SR 3.4.16.1 surveillance for RCS gross non-iodine specific activity. The surveillance provides an indication of any increase in the noble gas specific activity. The results of the surveillance on DEX allow proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

SR 3.4.16.1 is modified by inclusion of a NOTE which states, Only required to be performed in MODE 1. This NOTE modifies the SR to permit entry into the applicable MODE(S) before performing the surveillance. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation. This allows entry into MODE 4, MODE 3, and MODE 2 prior to performing the surveillance. This allows the surveillance to be performed in any of those MODES, prior to entering MODE 1, similar to the current surveillance SR 3.4.16.2 for DEI.

3.1.9 SR 3.4.16.3 Deletion The current SR 3.4.16.3 which required the determination of E Bar is deleted. TS 3.4.16 LCO on RCS specific activity supports the dose analyses for DBAs, in which the whole body dose is primarily dependent on the noble gas concentration, not the non-gaseous activity currently captured in the E Bar definition. With the elimination of the limit for RCS gross specific activity and the addition of the new LCO limit for noble gas specific activity, this SR to determine E Bar is no longer required.

3.2 Precedent The TS developed for the Westinghouse AP600 and AP1000 advanced reactor designs incorporate a limiting condition for operation (LCO) for RCS Dose Equivalent Xenon (DEX) activity in place of the LCO on non-iodine gross specific activity based on E Bar. This approach was approved by the NRC staff for the AP600 in NUREG-1512, Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003, dated August 1998 and for the AP1000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004. In addition, the curve describing the maximum allowable iodine concentration during the 48-hour period of elevated activity as a function of power level was not included in the TS approved for the AP600 and API000 advanced reactor designs.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York 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 installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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 (73 FR 15786). 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) 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: A. Jason Lising Date: June 17, 2008