ML081400035

Issuance of Amendment Technical Specification Task Force TSTF-490, Rev 0, Deletion of E Bar Definition and Revision to RCS Specific Activity.

ML081400035
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/08/2008
From:	Wang A NRC/NRR/ADRO/DORL/LPLIV
To:	Entergy Operations
Wang, A B, NRR/DORL/LPLIV, 415-1445
Shared Package
ML081400034	List: ML081400035 ML081400037
References
TAC MD8312
Download: ML081400035 (16)

Text

September 8, 2008 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE:

TECHNICAL SPECIFICATION TASK FORCE TSTF-490, REVISION 0, ADELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY (TAC NO. MD8312)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 282 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit No. 2 (ANO-2). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 13, 2008.

The amendment replaces the current ANO-2 TS 3.4.8, RCS [reactor coolant system] Specific Activity, limit on RCS gross specific activity with a new limit on RCS noble gas specific activity.

The noble gas specific activity limit would be based on a new dose equivalent Xe-133 (DEX) definition that would replace 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).

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

Sincerely,

/RA/

Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368

Enclosures:

1. Amendment No. 282 to NPF-6

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

Pkg ML081400034 (Amdt. ML081400035, License/TS Pgs ML081400037)

(*) with comments OFFICE NRR/LPL4/PM NRR/LPL4/LA DRA/AADB/BC DIRS/ITSB/BC OGC, NLO NRR/LPL4/BC NRR/LPL4/PM NAME AWang GLappert RTaylor RElliott (*) AGendelman MMarkley AWang DATE 5/29/08 5/29/08 5/30/08 6/18/08 8/21/08 9/8/08 9/8/08 Arkansas Nuclear One (2/25/08) cc:

Senior Vice President Pope County Judge Entergy Nuclear Operations Pope County Courthouse P.O. Box 31995 100 W. Main Street Jackson, MS 39286-1995 Russellville, AR 72801 Vice President, Oversight Senior Resident Inspector Entergy Nuclear Operations U.S. Nuclear Regulatory Commission P.O. Box 31995 P.O. Box 310 Jackson, MS 39286-1995 London, AR 72847 Senior Manager, Nuclear Safety Regional Administrator, Region IV

& Licensing U.S. Nuclear Regulatory Commission Entergy Nuclear Operations 611 Ryan Plaza Drive, Suite 400 P.O. Box 31995 Arlington, TX 76011-8064 Jackson, MS 39286-1995 Senior Vice President

& Chief Operating Officer Entergy Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Associate General Counsel Entergy Nuclear Operations P.O. Box 31995 Jackson, MS 39286-1995 Manager, Licensing Entergy Operations, Inc.

Arkansas Nuclear One 1448 SR 333 Russellville, AR 72802 Section Chief, Division of Health Radiation Control Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Section Chief, Division of Health Emergency Management Section Arkansas Department of Health and Human Services 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 282 Renewed License No. NPF-6

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

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated March 13, 2008, 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 license 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 Renewed Facility Operating License No. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-6 Technical Specifications Date of Issuance: September 8, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 282 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Renewed Facility Operating License No. NPF-6 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.

Operating License REMOVE INSERT Page Page Technical Specifications REMOVE INSERT 1-4 1-4 3/4 4-18 3/4 4-18

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 282 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT NO. 2 DOCKET NO. 50-368

1.0 INTRODUCTION

By letter dated March 13, 2008 to the U.S. Nuclear Regulatory Commission (NRC) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080850908), Entergy Operations, Inc. (Entergy, the licensee) proposed changes to the technical specifications (TS) for Arkansas Nuclear One, Unit No.2 (ANO-2). The requested changes are the adoption of Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, ADeletion of E Bar Definition and Revision to RCS [reactor coolant system] Specific Activity Technical Specification,@ for pressurized water reactor (PWR) Standard Technical Specifications (STS).

The changes were approved by the NRC staff Safety Evaluation (SE) dated September 27, 2006 (ADAMS Accession No. ML062700612). TSTF-490 was announced for availability in the Federal Register on March 8, 2007, as part of the consolidated line item improvement process (CLIIP). The TSTF involves changes to NUREG-1432, the STS for Combustion Engineering Plants Section like ANO-2, 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 ANO-2 equivalent TS to the STS 3.4.16 is TS 3.4.8, RCS [reactor coolant system] Specific Activity. In its application, the licensee stated the justifications presented in the TSTF-490, Revision 0, and the model SE prepared by the NRC are applicable to ANO-2 and justify the proposed amendment for ANO-2.

2.0 REGULATORY EVALUATION

Section 50.36 of Title of the Code of Federal Regulations (10 CFR 50.36), the Commission established its regulatory requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings;

(2) limiting conditions for operation (LCOs); (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls. The proposed changes affect safety limits, LCOs, and SRs.

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 technical specifications (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, AEC, 1962, ACalculation of Distance Factors for Power and Test Reactors Sites,@ in their dose consequence analyses, the NRC staff uses the regulatory guidance (RG) provided in NUREG-0800, AStandard 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 RG 1.195, AMethods 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.

The applicable dose criteria 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) Part 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 is 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, AMethods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light Water Nuclear Power Reactors,@ May 2003.

For control room dose consequence analyses that use the TID 14844 source term, the regulatory requirement for which the NRC staff bases its acceptance is General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50, AControl 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, AControl 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, AControl Room Habitability at Light Water Nuclear Power Reactors,@ May 2003, Section C.4.5 of RG 1.195, May 2003, states that in lieu of the dose equivalency guidelines from Section 6.4 of NUREG-0800, the 10 CFR 20.1201 annual organ dose limit of 50 rem can be used for both the thyroid and skin dose equivalent of 5 rem whole body.

3.0 TECHNICAL EVALUATION

3. 1 Technical evaluation of TSTF-490 TS changes 3.1.1 Revision to the Definition of DEI The licensee has adopted the standard wording from the STS for the definition of DEI. The list of acceptable DCFs for use in the determination of DEI include the following:

$ Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

$ Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

$ ICRP 30,1979, page 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity."

$ Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.

$ Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

Entergy states that the ANO-2 analyses employ Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites, to determine the DEI. As previously stated, it is incumbent on the licensee to ensure that the DCFs used in the determination of DEI are consistent with the DCFs used in the applicable dose consequence analyses. The change will allow the licensee to calculate DEI using the same DCFs as are used in the dose consequence analyses and is, therefore, acceptable.

3.1.2 Deletion of the Definition of E Bar and the Addition of a New Definition for Dose Equivelant Xe-133 (DEX)

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 would use either the average gamma disintegration energies for the nuclides or the effective dose conversion factors from Table III.1 of 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 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.

The change will implement a 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 3.4.8 specifies the limit for primary coolant gross specific activity as 100/E Bar µCi/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 3.4.8 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 - 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 the average gamma disintegration energies derived from the dose conversion factors provided in ICRP Publication 2, Report of ICRP Committee on Permissible Dose for Internal Radiation, for non-fuel damage events.

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. The licensee in its submittal has stated it has ensured that the DCFs used in the determination of DEI and the newly defined DEX are consistent with the DCFs used in the applicable dose consequence analyses.

3.1.3 LCO 3.4.8, ARCS Specific Activity@

LCO 3.4.8 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:

a. 1.0 µCi / gram DOSE EQUIVALENT I-131, and

b. 100/E µCi / gram.

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

The DEI limit of 1.0 Ci/gm as defined in current Action a is maintained in proposed SR 4.4.8.2.

The licensee states that the limit of 1.0 Ci/gm is consistent with the current SGTR and MSLB radiological consequence analyses.

3.1.4 TS 3.4.8 Applicability TS 3.4.8 Applicability is modified to exclude all of MODE 5. 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 and the proposed change to exclude MODE 5 is acceptable from a radiological dose perspective. This is consistent with the STS as modified by TSTF-490.

3.1.5 TS 3.4.8 ACTION a DEI TS 3.4.8 ACTION a is revised by :

a. Replacing the DEI site specific Action when A >1.0 µCi/gm@ and conditions with the words:

a. With DOSE EQUIVALENT I-131 not within limit:

1. Verify DOSE EQUIVALENT 1-131 60 µCi / gram once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and

2. Restore DOSE EQUIVALENT I-131 within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

The wording is consistent with the current TS 3.4.8.a and Table 4.4-4, Item 4.a and the revised STS 3.4.16 LCO format. The site specific DEI limit of 1.0 µCi/gm is unchanged and now contained in proposed SR 4.4.8.2. The Allowed Outage Time for revised TS 3.4.8 Action a.2 will require restoration of DEI to within limit in 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 current TS 3.4.8

Action a, Allowed Outage Time for DEI limit exceedence. A Note is also added to the revised Action a 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 3.4.8 is not being met and is the same Note that will be added for Action b. The proposed Note would allow entry into the applicable Modes from MODE 4 to MODE 1 (power operation) while the DEI limit is exceeded and the DEI is being restored to within its limit. This Mode change is acceptable due to the significant conservatism incorporated into the DEI specific activity limit, the low probability of an event occurring which is limiting due to exceeding the DEI specific activity limit, and the ability to restore transient specific excursions while the plant remains at, or proceeds to power operation.

The proposed format change will not alter current STS requirements and is acceptable from a radiological dose perspective.

In addition, TS 3.4.8 Action a is revised to remove the reference to Figure 3.4-1, "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 3.4-1 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.-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 3.4.8 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 3.4.8 ACTION b Revision to include Action for DEX Limit TS 3.4.8 Action b is replaced with a new Action b, DEX not within limits. This change is made to be consistent with the change to the STS 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 µCi/gm is contained in revised SR 4.4.8.1. The site specific limit of DEX in µCi/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 3.4.8 LCO on RCS specific activity and its associated Actions 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 Allowed Action Time for revised TS 3.4.8 Action b will require restoration of DEX to within limit in 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 current TS 3.4.8 Action a and the proposed Allowed Outage Time for TS 3.4.8 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 Allowed Outage Time for noble gas activity being out of specification in the revised Action b should be at least as great as the Allowed Outage Time for iodine specific activity being out of specification in current Action a. Therefore the Allowed Outage Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for revised Action b is acceptable from a radiological dose perspective. A Note is also added to the revised Action b 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 3.4.8 is not being met and is the same Note that is currently stated for the STS 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.

A Note is also added to the revised Action b 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 3.4.8 is not being met and is the same Note that is currently stated for Action a. 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 3.4.8 Action c TS 3.4.8 Action c is revised from:

c. With the specific activity of the primary coolant >1.0 µCi/gram DOSE EQUIVALENT I-131 or >100/E µCi/gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

to:

c. With the requirements of Action a and/or b not met, or with DOSE EQUIVALENT I-131 >60 µCi/gm, be in at least HOT STANDBY in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The proposed change is consistent with the changes made to TS 3.4.8 Action a and b which now provide the same completion time for both components of RCS specific activity as discussed in the revision to Action b. The only deviation is Entergy has used the noun name instead of the Mode number. The TS definitions define Mode 3 as Hot Standby and Mode 5 as Cold Shutdown so the proposed change is equivalent to the STS and acceptable. The revision to Action c also maintains the current limit on DEI with a site specific value of >60 µCi/gm. This change makes Action c consistent with the changes made to TS 3.4.8 Action a.1.

The change to TS 3.4.8 Action c requires the plant to be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 /> when the requirements of Action a and/or b are not me or with DOSE EQUIVALENT I-131 >60 µCi/gm. These changes are consistent with the changes made to the TS 3.4.8 Applicability and STS 3.4.16 Required Action C. 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.

As discussed above, the new Action has been added to TS 3.4.8 Action c of an Allowed Outage Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for the plant to reach MODE 5. The proposed change is consistent with STS 3.4.16 Required Action C. This Allowed Outage 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 Allowed Outage Time to reach MODE 5.

3.1.8 SR 4.4.8 Surveillance Requirements Current SR 4.4.8 states that the specific activity of the primary coolant shall be determined to be within limits by performance of the sampling and analysis program on TS Table 4.4-4. Entergy has proposed to replace SR 4.4.8 and remove Table 4.4-4 with the language consistent with STS SR 3.4.16.

3.1.9 New SR 4.4.8.1 DEX Surveillance The change replaces the current SR 4.4.8, Table 4.4-4, Item 1 for RCS gross specific activity with a surveillance to verify that the site specific reactor coolant DEX specific activity is 1200

µCi/gm. The change provides a surveillance for the new LCO limit added to TS 3.4.8 for DEX.

The new SR 4.4.8.1 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 less frequent than required under the current SR 4.4.8.1 surveillance for RCS gross non-iodine specific activity but is consistent with STS SR 3.4.16.1. 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 4.4.8.1 is modified by inclusion of a NOTE which state, Only required to be performed in MODE 1. 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 4.4.8.2 for DEI.

3.1.10 New SR 4.4.8.2 The change replaces the current SR 4.4.8, Table 4.4-4, Items 2 and 4 for DEI. The current DEI limit of 1.0 Ci/gm is retained in proposed SR 4.4.8.2. The licensee states that the limit of 1.0 Ci/gm is consistent with the current SGTR and MSLB radiological consequence analyses. The limits for DEI are discussed in Section 3.1.3 and 3.1.5. The new SR 4.4.8.2.a surveillance requires performing a gamma isotopic analysis for DEI of the reactor coolant at least once every

14 days, which is the same frequency required under the current SR 4.4.8, Table 4.4-4, Item 2 and is consistent with STS SR 3.4.16.2. The new SR 4.4.8.2.b surveillance requires performing a gamma isotopic analysis for DEI of the reactor coolant between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a thermal power change exceeding 15 percent of the rated thermal power within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, which is the same frequency required under the current SR 4.4.8. Table 4.4-4, Item 2.b and is consistent with STS SR 3.4.16.2. The results of the surveillance on DEI allow proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

SR 4.4.8.2 is modified by inclusion of a NOTE which state, Only required to be performed in MODE 1. The 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 4.4.8.1 for DEX.

3.1.11 Deletion of Table 4.4-4 Item 3, Radiochemical for Determination The current SR 4.4.8, Table 4.4-4, Item 3 which required the determination of E Bar is deleted.

TS 4.4.8 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 and Conclusion The technical specifications developed for the Westinghouse AP600 and AP1000 advanced reactor designs incorporate an LCO for RCS 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.

Based on the above evaluation, the NRC staff concludes the proposed changes meet the technical requirements in the regulations that are discussed in Section 2.0 of this safety evaluation. Based on this, the NRC staff further concludes that the proposed TS changes in the proposed amendment meet 10 CFR 50.36 and, therefore, the proposed amendment is acceptable.

The licensee identified changes to be made to the TS Bases that are associated with the TSs that are being changed in Attachment 4 to its application. The NRC does not approve these changes. The changes to the TS Bases to the TS Bases are made by the licensee through TS 5.5.14, Technical Specifications (TS) Bases Control Program. However, the NRC has reviewed the identified changes to the TS Bases for this amendment and does not have any problems with these changes.

4.0 STATE CONSULTATION

In accordance with the Commission=s regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment[s] change[s] a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or 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 published May 6, 2008 (73 FR 25039). 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: Alan B. Wang Date: September 8, 2008