Text

Response to Request for Additional Information Related to License Amendment Request to Adopt Traveler TSTF-490, Rev 0, Deletion of E Bar Definition & Revision to RCS Specific Activity
	ML11271A027
Person / Time
Site:	Oconee, Mcguire, Catawba, McGuire
Issue date:	09/22/2011
From:	Morris J; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TSTF-490, Rev 0
	Download: ML11271A027 (67)
	v • d • e

James R. Morris Duke Vice President, rEEnergy Catawba Nuclear Station Nuclear Generation Duke Energy Corporation Catawba Nuclear Station / CN01 VP 4800 Concord Road York, SC 29745 September 22, 2011 10 CFR 50.90 803-701-4251 803-701-3221 Jira. Morris@duke-energy.corn U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station (CNS), Units 1 and 2 CNS Docket Nos. 50-413, 50-414 McGuire Nuclear Station (MNS), Units 1 and 2 MNS Docket Nos. 50-369, 50-370 Oconee Nuclear Station (ONS), Units 1, 2, and 3 ONS Docket Nos. 50-269, 50-270, 50-287 Response to Request for Additional Information Related to License Amendment Request to Adopt Traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity"

REFERENCES:

1. Letter from Duke Energy, LLC to U.S. NRC, Technical Specifications Revision Request to Adopt TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity," dated December 15, 2009.

2. Federal Register, Volume 72, Number 50, Thursday, March 15, 2007, Page 12217, "Notice of Availability of Model Application Concerning Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process."

This letter provides a response to a Request for Additional Information (RAI) regarding the Duke Energy License Amendment Request (LAR) dated December 15, 2009 related to TSTF-490, Revision 0. The request was conveyed by the NRC staff, John F. Stang via email correspondence on June 22, 2011.

The NRC staff's questions and Duke Energy's responses are provided in the Enclosure.

www.duke-energy.com

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U.S. Nuclear Regulatory Commission September 22, 2011 Page 2 The original determination that the LAR contains No Significant Hazards Considerations and the basis for the categorical exclusion of performing an Environmental/Impact Statement has not changed as a result of this RAI.

In addition, the Duke Energy Fleet requests a 120 day implementation upon issuance of the subject amendment in lieu of the previously requested 60 day implementation grace period. This request is based upon the parallel implementation of TSTF-425 at ONS and to allot time to make the necessary Updated Final Safety Analysis Review (UFSAR) changes required.

Final Technical Specification (TS) and Bases pages have been provided as Attachments. They were not included in the initial submittal due to current amendments pending NRC review that would directly impact the issuance of the final TS pages within this submittal.

Pursuant to 10 CFR 50.91, a copy of the RAI responses is being sent to the appropriate state officials.

This letter contains no new commitments as a result of this RAI.

Please contact A.F. Driver at 803-701-3445 or Adrienne.Driver(,duke-energy.com.

Sincerely,

Enclosure:

Request For Additional Information Related to the Adoption of TSTF-490, Rev.0, LAR dated December 15, 2009 CNS, MNS, ONS Technical Specification Mark-Ups Pages CNS, MNS, ONS Technical Specification and Bases Final Pages

U.S. Nuclear Regulatory Commission September 22, 2011 Page 3 Oath or Affirmation James R. Morris affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

Site Vicepresident, Catawba Nuclear Station Subscribed and sworn to me:

, - /

Date NotaryPb F

My Commission Expires:

Date SEAL

U.S. Nuclear Regulatory Commission September 22, 2011 Page 4 xc:

V. M. McCree, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, Georgia 30303-1257 J. F. Stang, Jr., Senior Project Manager (ONS)

U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 G9A Rockville, MD 20852-2738 J. H. Thompson, Project Manager (CNS & MNS)

U. S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8 G9A Rockville, MD 20852-2738 A. T. Sabisch NRC Senior Resident Inspector Oconee Nuclear Station John Zeiler NRC Senior Resident Inspector McGuire Nuclear Station G. A. Hutto, III NRC Senior Resident Inspector Catawba Nuclear Station S. E. Jenkins, Manager Radioactive & Infectious Waste Management SC Dept. of Health and Env. Control 2600 Bull St.

Columbia, SC 29201 W. L. Cox, III, Section Chief Div. of Environmental Health, RP Section NC Dept. of Env. & Natural Resources 1645 Mail Service Center Raleigh, NC 27699-1645

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 NRC Question 1:

1. The Federal Register Notice for the "Availability of the Model Application for TSTF-490" (72 Fed. Reg. 12217) states:

To efficiently process the incoming license amendment applications, the NRC staff requests that each licensee applying for the changes addressed by TSTF-490, Revision 0, using the CLIIP submit a license amendment request (LAR) that adheres to the following model. Any variations from the model LAR should be explained in the licensee's submittal. Variations from the approach recommended in this notice may require additional review by the NRC staff, and may increase the time and resources needed for the review. Significant variations from the approach, or inclusion of additional changes to the license, will result in staff rejection of the submittal. Instead, licensees desiring significant variations and/or additional changes should submit a LAR that does not claim to adopt TSTF-490.

The NRC model application (ML070250176) provides the expected level of details and content to adopt TSTF-490, Revision 0. The NRC model includes three enclosures.

These enclosures are: 1) a description and assessment of proposed changes, 2) the proposed Technical Specification changes and Technical Specification bases changes and 3) the final Technical Specification and Bases pages. The model LAR specified that the information contained in these enclosures is needed for the review of the LAR. Duke provided the information identified in Enclosures 1 and 2 of the model application, but did not provide the information from Enclosure 3. In addition, the LAR does not contain an explanation for this variation from the model. Consistent with TSTF-490 and the CLIIP please provide the information from Enclosure 3 or provide an explanation for why the information is not needed to support the review.

Duke Energy Response 1:

The final Technical Specification and Bases pages were not provided in the initial submittal consistent with normal practices for a License Amendment Request submitted to the NRC on behalf of Duke Energy Carolinas, LLC. The Regulatory Compliance staff provides the final Technical Specification and Bases pages prior to the issuance of the amendment. This is communicated and provided to the respective NRR Project Manager. Final Technical Specification pages were not provided with the initial submittal since, at the time of the initial submittal, issuance of other LARs which were currently under NRC review were expected to impact one another. This is not a variation from the approach for adoption of TSTF-490, Rev. 0. The final TS and Bases pages have been included as attachments within this submittal (Reference Attachment 2).

Enclosure Page 1 Enclosure Page 1

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 NRC Question 2:

2. In Section 2.0, "Proposed Changes," of the subject license amendment request, the licensee proposed TS changes to revise Limiting Condition for Operation (LCO) 3.4.16, "RCS Specific Activity," for both Catawba and McGuire, and LCO 3.4.11, "RCS Specific Activity," for Oconee, APPLICABILITY requirements to specify that the LCO is applicable in MODES 1, 2, 3, and 4. In accordance with this proposal, the licensee also proposed to add the NOTE that states, "Only required to be performed in MODE 1." to the surveillance requirements (SR) of the TS, thus removing the applicability of the surveillance requirements to other MODES.

The NRC staff has a concern about the proposed addition of the aforementioned NOTE.

The proposed change revises the conditions for sampling, and may exclude sampling during the plant conditions where LCO 3.4.16 may be exceeded. After transient conditions (e.g. reactor trip, plant depressurization, shutdown or startup) that end in MODES 2, 3, or 4, the SR is not required to be performed. Isotopic spiking and fuel failures are more likely during transient conditions than during steady state plant operations.

Because LCO 3.4.16 for Catawba and McGuire, and LCO 3.4.11 for Oconee could potentially be exceeded after plant transient or power changes, please justify why sampling is no longer needed in the plant MODES that are proposed to be eliminated and justify how the LCO 3.4.16 for Catawba and McGuire, and LCO 3.4.11 for Oconee remains consistent with the design bases analysis from which the LCO limits are derived (ie. main steamline break, steam generator tube rupture, etc.). Furthermore, please justify why there is an apparent disparity between the modes of applicability (MODES 1, 2, 3, and 4) and the limited mode (MODE 1) under which the surveillance is required.

In addition, please verify that each facility can perform the required surveillance for both DEI and DEX in all MODES of applicability (MODES 1, 2, 3, and 4). If not, please justify why performing the required surveillance in any of aforementioned MODES is not needed to ensure the safe operation of the facility, in accordance with the licensee's current licensing bases.

Please provide a response addressing each individual facility (e.g. Catawba, McGuire, and Oconee).

Enclosure Page 2 Enclosure Page 2

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 Duke Energy Response 2:

CNS Response:

2.a.

Justify why there is an apparent disparity between the modes of applicability (MODES 1, 2, 3, and 4) and the limited mode (MODE 1) under which the surveillance is required. Please justify why sampling is no longer needed in the plant MODES that are proposed to be eliminated.

With this supplement, Duke Energy proposes changing the SR 3.4.16.1 Note associated with DEX to state "Only required to be performed in MODES 1, 2, and 3 with Reactor Coolant System (RCS) average temperature ? 500 OF." Duke Energy also proposes deleting the SR 3.4.16.2 Note associated with DEI in its entirety. These changes from the original LAR submittal reduce the apparent disparity between the LCO modes of applicability and modes under which the surveillance requirements are required. The changes proposed per this supplement are a conservative and more restrictive deviation from TSTF-490, Revision 0. It is also important to the note that the revised SR 3.4.16.1 Note for DEX proposed with this supplement is consistent with the LCO modes of applicability of the current TS 3.4.16. Furthermore, the deletion of SR 3.4.16.2 Note for DEI in its entirety provides continued assessment of RCS activity for all modes of applicability since DEI will no longer be limited to only MODE 1 operation as is the case in the current TS. Given that iodine is the dominant contributor in Catawba Nuclear Station (CNS) Steam Generator Tube Rupture (SGTR) and Main Steam Line Break (MSLB) dose analysis and that CNS is proposing to sample DEI down through MODE 4, the small disparity between the modes of applicability and the modes that require sampling under SR 3.4.16.1 for DEX is insignificant (Please reference 2c for additional detail).

2.b.

Justify how the LCO 3.4.16 for Catawba remains consistent with the design bases analysis from which the LCO limits are derived (ie. main steamline break, steam generator tube rupture, etc.).

The proposed changes to LCO 3.4.16 for Catawba remain consistent with the CNS design basis analysis from which the LCO limits are derived (i.e., MSLB and SGTR).

The current TS 3.4.16 LCO applicability requirements (prior to this proposed change) are limited to Modes 1, 2 and Mode 3 with RCS average temperature > 500 OF. The CNS Design Basis (DB) SGTR and MSLB analyses are based on power operating (Mode 1) conditions and assume RCS activity at TS limits (please refer to the response to RAI

3). Therefore, these design basis analyses are bounding for the lower modes of operation. Furthermore, while a SGTR or MSLB is possible in the lower modes of operation, the likelihood is low particularly considering operation in Mode 4 is temporary and would not normally exceed the sample frequency required in the surveillance requirement.

Enclosure Page 3 Enclosure Page 3

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 2.c Verify that each facility can perform the required surveillance for both DEI and DEX in all MODES of applicability (MODES 1, 2, 3, and 4). If not, please justify why performing the required surveillance in any of aforementioned MODES is not needed to ensure the safe operation of the facility, in accordance with the licensee's current licensing bases.

The proposed SR 3.4.16.1 Note restricting the sampling of DEX to MODES 1, 2, and 3 with RCS average temperature > 500°F is needed since RCS pressure outside these modes may be insufficient to ensure collection of a representative, homogeneous sample for analysis of gaseous isotopes. Restricting the surveillance to Modes 1, 2, and 3 with Tave > 500°F ensures the necessary plant conditions have been established to produce sufficient pressure and sample flow.

The need for a DEX sample is not significant in the lower modes of operation since the pressures and temperatures on the steam generator tubes required to initiate or increase the consequence of a SGTR or MSLB are reduced. Steam generator tubes are designed against thermodynamic and other stresses placed upon them at full power operations and accident conditions. The TS RCS leakage limit restricts the amount of leakage that may occur through the steam generators. The RCS activity TS is intended to limit the amount of activity that could be released to the public should a SGTR occur.

Since the conditions required to initiate or increase the consequence of a SGTR or MSLB are reduced in the lower modes and SG leakage is limited by the RCS TS, the significance of performing samples to determine DEX activity in RCS below 500 OF is negligible and therefore sampling in the plant MODES proposed to be eliminated for the DEX surveillance requirement is not necessary.

In addition to the above, Emergency Operating Procedures require Operators to rapidly reduce RCS pressure in any SG tube leak event that requires a unit shutdown (this is administratively required for SG tube leakage well below TS limits). The goal of the procedures is to reduce RCS pressure to be equal to or slightly less than SG pressure so that RCS fluid will cease to enter the secondary side of the SG through the leaking tube. These actions aid in preventing or limiting any release to the environment, regardless of RCS activity levels.

The above changes have no adverse impact to nuclear or public safety since the adoption of these proposed changes results in a TS that is more conservative in regard to the existing requirements. Therefore, the DEI and DEX surveillance requirements are in accordance with the licensee's current licensing basis and the DEX surveillance requirement is not needed in the plant modes proposed to be eliminated to ensure the safe operation of the facility.

MNS Response:

2.a.

Justify why there is an apparent disparity between the modes of applicability (MODES 1, 2, 3, and 4) and the limited mode (MODE 1) under which the surveillance is required. Please justify why sampling is no longer needed in the plant MODES that are proposed to be eliminated.

Enclosure Page 4

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 With this supplement, Duke Energy proposes changing the SR 3.4.16.1 Note associated with DEX to state "Only required to be performed in MODES 1, 2, and 3 with RCS average temperature a 500 'F." Duke Energy also proposes deleting the SR 3.4.16.2 Note associated with DEI in its entirety. These changes from the original LAR submittal reduce the apparent disparity between the LCO modes of applicability and modes under which the surveillance requirements are required. The changes proposed per this supplement are a conservative and more restrictive deviation from TSTF-490, Revision

0. It is also important to the note that the revised SR 3.4.16.1 Note for DEX proposed with this supplement is consistent with the LCO modes of applicability of the current TS 3.4.16. Furthermore, the deletion of SR 3.4.16.2 Note for DEI in its entirety provides continued assessment of RCS activity for all modes of applicability since DEI will no longer be limited to only MODE 1 operation as is the case in the current TS. Given that iodine is the dominant contributor in McGuire Nuclear Station (MNS) SGTR and MSLB dose analysis and that MNS is proposing to sample DEI down through MODE 4, the small disparity between the modes of applicability and the modes that require sampling under SR 3.4.16.1 for DEX is insignificant (Please reference 2c for additional detail).

2.b.

Justify how the LCO 3.4.16 for McGuire remains consistent with the design bases analysis from which the LCO limits are derived (ie. main steamline break, steam generator tube rupture, etc.).

The proposed changes to LCO 3.4.16 for McGuire remain consistent with the MNS design basis analysis from which the LCO limits are derived (i.e., MSLB and SGTR).

The current TS 3.4.16 LCO applicability requirements (prior to this proposed change) are limited to Modes 1, 2 and Mode 3 with RCS average temperature > 500 *F. The MNS DB SGTR and MSLB analyses are based on power operating (Mode 1) conditions and assume RCS activity at TS limits (please refer to the response to RAI #3). Therefore, these design basis analyses are bounding for the lower modes of operation.

Furthermore, while a SGTR or MSLB is possible in the lower modes of operation, the likelihood is low particularly considering operation in Mode 4 is temporary and would not normally exceed the sample frequency required in the surveillance requirement.

2.c Verify that each facility can perform the required surveillance for both DEI and DEX in all MODES of applicability (MODES 1, 2, 3, and 4). If not, please justify why performing the required surveillance in any of aforementioned MODES is not needed to ensure the safe operation of the facility, in accordance with the licensee's current licensing bases.

The proposed SR 3.4.16.1 Note restricting the sampling of DEX to MODES 1, 2, and 3 with RCS average temperature a 500°F is needed since RCS pressure outside these modes may be insufficient to ensure collection of a representative, homogeneous sample for analysis of gaseous isotopes. Restricting the surveillance to Modes 1, 2, and 3 with Tave a 500°F ensures the necessary plant conditions have been established to produce sufficient pressure and sample flow.

Enclosure Page 5 Enclosure Page 5

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 The need for a DEX sample is not significant in the lower modes of operation since the pressures and temperatures on the steam generator tubes required to initiate or increase the consequence of a SGTR or MSLB are reduced. Steam generator tubes are designed against thermodynamic and other stresses placed upon them at full power operations and accident conditions. The TS RCS leakage limit restricts the amount of leakage that may occur through the steam generators. The RCS activity TS is intended to limit the amount of activity that could be released to the public should a SGTR occur.

Since the conditions required to initiate or increase the consequence of a SGTR or MSLB are reduced in the lower modes and SG leakage is limited by the RCS TS, the significance of performing samples to determine DEX activity in RCS below 500 OF is negligible and therefore sampling in the plant MODES proposed to be eliminated for the DEX surveillance requirement is not necessary.

In addition to the above, Emergency Operating Procedures require Operators to rapidly reduce RCS pressure in any SG tube leak event that requires a unit shutdown (this is administratively required for SG tube leakage well below TS limits). The goal of the procedures is to reduce RCS pressure to be equal to or slightly less than SG pressure so that RCS fluid will cease to enter the secondary side of the SG through the leaking tube. These actions aid in preventing or limiting any release to the environment, regardless of RCS activity levels.

The above changes have no adverse impact to nuclear or public safety since the adoption of these proposed changes results in a TS that is more conservative in regard to the existing requirements. Therefore, the DEI and DEX surveillance requirements are in accordance with the licensee's current licensing basis and the DEX surveillance requirement is not needed in the plant modes proposed to be eliminated to ensure the safe operation of the facility.

ONS Response:

2.a.

Justify why there is an apparent disparity between the modes of applicability (MODES 1, 2, 3, and 4) and the limited mode (MODE 1) under which the surveillance is required. Please justify why sampling is no longer needed in the plant MODES that are proposed to be eliminated.

With this supplement, Duke Energy proposes changing the SR 3.4.11.1 Note associated with DEX to state "Only required to be performed in MODES 1, 2, and 3 with RCS average temperature > 500 OF." Duke Energy also proposes deleting the SR 3.4.11.2 Note associated with DEI in its entirety. These changes from the original LAR submittal reduce the apparent disparity between the LCO modes of applicability and modes under which the surveillance requirements are required. The changes proposed per this supplement are a conservative and more restrictive deviation from TSTF-490, Revision

0. It is also important to the note that the revised SR 3.4.11.1 Note for DEX proposed with this supplement is consistent with the LCO modes of applicability of the current TS 3.4.11. Furthermore, the deletion of SR 3.4.11.2 Note for DEI in its entirety provides continued assessment of RCS activity for all modes of applicability since DEI will no longer be limited to only MODE 1 operation as is the case in the current TS. Given that Enclosure Page 6 Enclosure Page 6

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 iodine is the dominant contributor in Oconee Nuclear Station (ONS) SGTR and MSLB dose analysis and that ONS is proposing to sample DEI down through MODE 4, the small disparity between the modes of applicability and the modes that require sampling under SR 3.4.11.1 for DEX is insignificant (Please reference 2c for additional detail).

2.b.

Justify how the LCO 3.4.11 for Oconee remains consistent with the design bases analysis from which the LCO limits are derived (ie. main steamline break, steam generator tube rupture, etc.).

The proposed changes to LCO 3.4.11 for Oconee remain consistent with the ONS design basis analysis from which the LCO limits are derived (i.e., MSLB and SGTR).

The current TS 3.4.11 LCO applicability requirements (prior to this proposed change) are limited to Modes 1, 2 and Mode 3 with RCS average temperature -> 500 OF. The ONS DB SGTR and MSLB analyses are based on power operating (Mode 1) conditions and assume RCS activity at TS limits (please refer to the response to RAI #3). Therefore, these design basis analyses are bounding for the lower modes of operation.

Furthermore, while a SGTR or MSLB is possible in the lower modes of operation, the likelihood is low particularly considering operation in Mode 4 is temporary and would not normally exceed the sample frequency required in the surveillance requirement.

2.c Verify that each facility can perform the required surveillance for both DEI and DEX in all MODES of applicability (MODES 1, 2, 3, and 4). If not, please justify why performing the required surveillance in any of aforementioned MODES is not needed to ensure the safe operation of the facility, in accordance with the licensee's current licensing bases.

The proposed SR 3.4.11.1 Note restricting the sampling of DEX to MODES 1, 2, and 3 with RCS average temperature -> 500°F is needed since RCS pressure outside these modes may be insufficient to ensure collection of a representative, homogeneous sample for analysis of gaseous isotopes. Restricting the surveillance to Modes 1, 2, and 3 with Tave -> 500°F ensures the necessary plant conditions have been established to produce sufficient pressure and sample flow.

The need for a DEX sample is not significant in the lower modes of operation since the pressures and temperatures on the steam generator tubes required to initiate or increase the consequence of a SGTR or MSLB are reduced. Steam generator tubes are designed against thermodynamic and other stresses placed upon them at full power operations and accident conditions. The TS RCS leakage limit restricts the amount of leakage that may occur through the steam generators. The RCS activity TS is intended to limit the amount of activity that could be released to the public should a SGTR occur.

Since the conditions required to initiate or increase the consequence of a SGTR or MSLB are reduced in the lower modes and SG leakage is limited by the RCS TS, the significance of performing samples to determine DEX activity in RCS below 500 °F is negligible and therefore sampling in the plant MODES proposed to be eliminated for the DEX surveillance requirement is not necessary.

In addition to the above, Emergency Operating Procedures require Operators to rapidly reduce RCS pressure in any SG tube leak event that requires a unit shutdown (this is administratively required for SG tube leakage well below TS limits). The goal of the Enclosure Page 7

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 procedures is to reduce RCS pressure to be equal to or slightly less than SG pressure so that RCS fluid will cease to enter the secondary side of the SG through the leaking tube. These actions aid in preventing or limiting any release to the environment, regardless of RCS activity levels.

The above changes have no adverse impact to nuclear or public safety since the adoption of these proposed changes results in a TS that is more conservative in regard to the existing requirements. Therefore, the DEI and DEX surveillance requirements are in accordance with the licensee's current licensing basis and the DEX surveillance requirement is not needed in the plant modes proposed to be eliminated to ensure the safe operation of the facility.

NRC Question 3:

Consistent with the model safety evaluation for TSTF-490, please confirm that the definitions and site-specific limits for both DEl and DEX, and the dose conversion factors (DCFs) used for the determination of DEI and DEX surveillances, are consistent with the current design basis radiological dose consequence analyses (i.e. SGTR and MSLB).

Also for both DEl and DEX, please provide the information necessary (dose conversion factors and reactor coolant system radioisotopic concentrations) for the staff to verify the proposed values in the LCO.

Please provide a response addressing each individual facility (e.g. Catawba, McGuire, and Oconee).

Duke Energy Response 3:

CNS Response The current design basis radiological dose consequence analyses for the CNS steam generator tube rupture (SGTR) and main steam line break (MSLB) accidents do not postulate fuel damage. These accidents are analyzed using the maximum reactor coolant system activity allowed by TS. The activity for these accidents is presented in the form of DEI and noble gas activity. The dose conversion factors associated with the determination of DEI used in these analyses are consistent with the CNS TS limits for DEI and for the determination of DEI surveillances requirements. As stated in the original TSTF-490 submittal dated December 15, 2009, all units at CNS have received license amendments for full implementation of Alternative Source Term (AST) methodology, pursuant to 10 CFR 50.67. Therefore, the DCFs used to determine DEI are from Table 2.1 of EPA Federal Guidance Report (FGR) No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." The DCFs for the determination for DEX are from EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil."

Enclosure Page 8 Enclosure Page 8

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 As stated in the proposed TS definition, DEX shall be that concentration of Xe-1 33 (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-1 31m, Xe-1 33m, Xe-133, Xe-135m, Xe-1 35, and Xe-1 38 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. As previously stated, the determination of DEX shall be performed using effective dose conversion factors for air submersion listed from Table 111.1 of EPA Federal Guidance Report No. 12.

The DEX is determined by multiplying each noble gas RCS concentration (pCi/gm) by its respective DCF from FGR12. Each of these values are then divided by the FGR12 DCF for Xe-133 and then summed to determine the DEX. The table below illustrates the calculation of DEX for the CNS design basis DEI accidents. The first column is the list of noble gas isotopes as defined in the proposed TS definition of DEX. Column two is the concentration of noble gases in pCi/gm assumed in for both the CNS SGTR and MSLB accidents. Column three lists the respective DCF for each noble gas isotope as given in FGR12, Table 111.1. The last column illustrates the calculation of DEX as previously described.

Calculation of CNS DEX Concentration FGR12, Table II1.1 DEX Isotope (PCi/gm)

DCFs (Sv-s/Bq-(pCi/gm) m3)

(__igm)

KR-85M 2.06E+00 7.48E-15 9.88E+00 KR-85 7.52E+00 1.19E-16 5.74E-01 KR-87 1.34E+00 4.12E-1 4 3.54E+01 KR-88 3.71 E+00 1.02E-1 3 2.43E+02 XE-1 31 M 2.27E+00 3.89E-16 5.65E-01 XE-133M 1.75E+01 1.37E-15 1.54E+01 XE-133 2.78E+02 1.56E-15 2.78E+02 XE-135M 4.95E-01 2.04E-14 6.47E+00 XE-135 7.42E+00 1.19E-14 5.66E+01 XE-138 6.59E-01 5.77E-14 2.44E+01 DEX I 6.70E+02 I As shown in the table above, the calculated DEX 6.70E+02 pCi/gm for the current CNS SGTR and MSLB DB radiological dose consequences analyses exceeds the proposed DEX limit of 280 pCi/gm. This provides a small degree of margin in the dose analyses available for future use.

As stated in the proposed TS definition, DEI shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same acute dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. As previously s tated, the determination of DEI shall be performed using the CDE or CEDE DCFs from Table 2.1 of EPA Federal Guidance Report No. 11.

Enclosure Page 9 Enclosure Page 9

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 The DEI is determined by multiplying each iodine RCS concentration (pCi/gm) by its respective DCF from FGR1 1. Each of these values is then divided by the FGR1 1 DCF for 1-131 and then summed to determine the DEI. The table below illustrates the calculation of DEI for the CNS design basis DEI accidents. The first column is the list of iodine isotopes as defined in the proposed TS definition of DEL. Column two is the concentration of iodine isotopes in pCi/gm assumed in for both the CNS SGTR and MSLB accidents. Column three lists the respective DCF for each iodine isotope as given in FGR1 1, Table 2.1. The last column illustrates the calculation of DEI as previously described.

Calculation of CNS DEI Isotope Concentration FGR11, Table 2.1 DEI (pCilgm)

DCFs (Sv/Bq)

(pCi/gm) 1-131 7.56E-01 8.89E-09 7.56E-01 1-132 2.72E-01 1.03E-10 3.15E-03 1-133 1.21 E+00 1.58E-09 2.15E-01 1-134 1.81 E-01 3.55E-11 7.25E-04 1-135 6.65E-01 3.32E-10 2.49E-02 DEI I 1.OOE+00 As shown in the table above, the calculated DEI for all units at CNS based on the current SGTR and MSLB DB radiological dose consequences analyses meets the DEI limit of 1.0 pCi/gm.

MNS Response:

The current design basis radiological dose consequence analyses for the MNS steam generator tube rupture (SGTR) and main steam line break (MSLB) accidents do not postulate fuel damage. These accidents are analyzed using the maximum reactor coolant system activity allowed by TS. The activity for these accidents is presented in the form of DEI and noble gas activity. The dose conversion factors associated with the determination of DEI used in these analyses are consistent with the MNS TS limits for DEI and for the determination of DEI surveillances requirements. As stated in the original TSTF-490 submittal dated December 15, 2009, all units at MNS have received license amendments for full implementation of Alternative Source Term (AST) methodology, pursuant to 10 CFR 50.67. Therefore, the DCFs used to determine DEI are from Table 2.1 of EPA Federal Guidance Report (FGR) No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." The DCFs for the determination for DEX are from EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil."

As stated in the proposed TS definition, DEX shall be that concentration of Xe-1 33 (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-1 33m, Xe-1 33, Xe-1 35m, Xe-1 35, and Xe-1 38 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. As previously stated, the determination of DEX shall be performed Enclosure Page 10 Enclosure Page 10

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 using effective dose conversion factors for air submersion listed from Table 111.1 of EPA Federal Guidance Report No. 12.

The DEX is determined by multiplying each noble gas RCS concentration (pCi/gm) by its respective DCF from FGR1 2. Each of these values are then divided by the FGR1 2 DCF for Xe-133 and then summed to determine the DEX. The table below illustrates the calculation of DEX for the MNS design basis DEI accidents. The first column is the list of noble gas isotopes as defined in the proposed TS definition of DEX. Column two is the concentration of noble gases in pCi/gm assumed in for both the MNS SGTR and MSLB accidents. Column three lists the respective DCF for each noble gas isotope as given in FGR12, Table 111.1. The last column illustrates the calculation of DEX as previously described.

Calculation of MNS DEX Concentration FGR12, Table 111.1 DEX Isotope (pCi/gm)

DCFs (Sv-s/Bq-(PCilgm) m3)

KR-85M 2.1OE+00 7.48E-15 1.01E+01 KR-85 8.80E+00 1.19E-16 6.71E-01 KR-87 1.20E+00 4.12E-14 3.17E+01 KR-88 3.70E+00 1.02E-13 2.42E+02 XE-131M 1.90E+00 3.89E-16 4.74E-01 XE-133M 3.1 OE+00 1.37E-15 2.72E+00 XE-133 2.81E+02 1.56E-15 2.81E+02 XE-135M 7.O0E-01 2.04E-14 9.15E+00 XE-135 6.30E+00 1.19E-14 4.81E+01 XE-138 7.OOE-01 5.77E-14 2.59E+01 DEX

[ 6.52E+02 As shown in the table above, the calculated DEX 6.52E+02 pCi/gm for the current MNS SGTR and MSLB DB radiological dose consequences analyses exceeds the proposed DEX limit of 280 pCi/gm. This provides a small degree of margin in the dose analyses available for future use.

As stated in the proposed TS definition, DEI shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same acute dose when inhaled as the combined activities of iodine isotopes I-131, 1-132, 1-133, 1-134, and 1-135 actually present. As previously stated the determination of DEI shall be performed using the CDE or CEDE DCFs from Table 2.1 of EPA Federal Guidance Report No. 11.

The DEl is determined by multiplying each iodine RCS concentration (pCi/g) by its respective DCF from FGR1 1. Each of these values is then divided by the FGR1 1 DCF for 1-131 and then summed to determine the DEl. The table below illustrates the calculation of DEI for the MNS design basis DEI accidents. The first column is the list of iodine isotopes as defined in the proposed TS definition of DEl. Column two is the concentration of iodine isotopes in pCi/gm assumed in for both the MNS SGTR and MSLB accidents. Column three lists the respective DCF for each iodine isotope as given Enclosure Page 11 Enclosure Page 11

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 in FGR1 1, Table 2.1. The last column illustrates the calculation of DEI as previously described.

Calculation of MNS DEI Concentration FGR11, Table 2.1 DEI Isotope (pCi/gm)

DCFs (Sv/Bq)

(pCi/gm) 1-131 7.56E-01 8.89E-09 7.56 E-01 1-132 2.72E-01 1.03E-10 3.15E-03 1-133 1.21 E+00 1.58E-09 2.15E-01 1-134 1.81 E-01 3.55E-11 7.25E-04 1-135 6.65E-01 3.32E-10 2.49E-02 DEI 1.OOE+O0 As shown in the table above, the calculated DEI for all units at MNS based on the current SGTR and MSLB DB radiological dose consequences analyses meets the DEI limit of 1.0 pCi/gm.

ONS Response:

The current design basis radiological dose consequence analyses for the ONS steam generator tube rupture (SGTR) and main steam line break (MSLB) accidents do not postulate fuel damage. These accidents are analyzed using the maximum reactor coolant system activity allowed by TS. The activity for these accidents is presented in the form of DEI and noble gas activity. The dose conversion factors associated with the determination of DEI used in these analyses are consistent with the ONS TS limits for DEI and for the determination of DEI surveillances requirements. As stated in the original TSTF-490 submittal dated December 15, 2009, all units at ONS have received license amendments for full implementation of Alternative Source Term (AST) methodology, pursuant to 10 CFR 50.67. Therefore, the DCFs used to determine DEI are from Table 2.1 of EPA Federal Guidance Report (FGR) No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." The DCFs for the determination for DEX are from EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil."

As stated in the proposed TS definition, DEX shall be that concentration of Xe-1 33 (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-1 33m, Xe-1 33, Xe-1 35m, Xe-1 35, and Xe-1 38 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. As previously stated, the determination of DEX shall be performed using effective dose conversion factors for air submersion listed from Table 111.1 of EPA Federal Guidance Report No. 12.

The DEX is determined by multiplying each noble gas RCS concentration (pCi/gm) by its respective DCF from FGR12. Each of these values are then divided by the FGR12 DCF for Xe-1 33 and then summed to determine the DEX. The table below illustrates the calculation of DEX for the ONS design basis DEI accidents. The first column is the list of Enclosure Page 12 Enclosure Page 12

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 noble gas isotopes as defined in the proposed TS definition of DEX. Column two is the concentration of noble gases in pCi/gm assumed in for both the ONS SGTR and MSLB accidents. Column three lists the respective DCF for each noble gas isotope as given in FGR12, Table 111.1. The last column illustrates the calculation of DEX as previously described.

Calculation of ONS DEX Concentration FGR12, Table II1.1 DEX Isotope (PCi/gm)

DCFs (Sv-s/Bq-(Pcilgm) m3)

KR-85M 2.15E+00 7.48E-15 1.03E+01 KR-85 1.82E+01 1.19E-16 1.39E+00 KR-87 1.18E+00 4.12E-14 3.11E+01 KR-88 3.69E+00 1.02E-13 2.41E+02 XE-131M 4.39E+00 3.89E-1 6 1.09E+00 XE-133M 5.76E+00 1.37E-15 5.06E+00 XE-133 3.93E+02 1.56E-15 3.93E+02 XE-135M 4.47E-01 2.04E-14 5.84E+00 XE-135 1.14E+01 1.19E-14 8.67E+01 XE-138 7.19E-01 5.77E-14 2.66E+01 I

DEX I 8.02E+02 As shown in the table above, the calculated DEX 8.02E+02 pCi/gm for the current ONS SGTR and MSLB DB radiological dose consequences analyses exceeds the proposed DEX limit of 280 pCi/gm. This provides a small degree of margin in the dose analyses available for future use.

As stated in the proposed TS definition, DEl shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same acute dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. As previously stated the determination of DEI shall be performed using the CDE or CEDE DCFs from Table 2.1 of EPA Federal Guidance Report No. 11.

The DEl is determined by multiplying each iodine RCS concentration (pCi/gm) by its respective DCF from FGR1 1. Each of these values is then divided by the FGR1i1 DCF for 1-131 and then summed to determine the DEl. The table below illustrates the calculation of DEl for the ONS design basis DEl accidents. The first column is the list of iodine isotopes as defined in the proposed TS definition of DEl. Column two is the concentration of iodine isotopes in pCi/gm assumed in for both the ONS SGTR and MSLB accidents. Column three lists the respective DCF for each iodine isotope as given in FGR1 1, Table 2.1. The last column illustrates the calculation of DEI as previously described.

Enclosure Page 13 Enclosure Page 13

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE ADOPTION OF TSTF-490, REV. 0 LAR, DATED DECEMBER 15, 2009 Calculation of ONS DEI Concentration FGRII, Table 2.1 DEI Isotope (pCi/gm)

DCFs (Sv/Bq)

(pCi/gm) 1-131 9.55E-01 8.89E-09 9.55E-01 1-132 1.42E-01 1.03E-10 1.64E-03 1-133 2.48E-01 1.58E-09 4.40E-02 1-134 1.32E-02 3.55E-11 5.27E-05 1-135 7.38E-02 3.32E-10 2.76E-03 DEI 1.OOE+00 As shown in the table above, the calculated DEI for all units at ONS based on the current SGTR and MSLB DB radiological dose consequences analyses meets the DEI limit of 1.0 pCi/gm.

Enclosure Page 14 Enclosure Page 14

Attachment I CNS, MNS, ONS Technical Specification Mark-Ups Pages

Definitions 1.1 1.1 Definitions (continued)

I!

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

DOSE EQU IVALENT 1131 shall b

e that hnnration o 1 13f 1microcurieprgam) that alone would produce the same thyold dosc as the suamtih' and whentpc inhalae eo 1431-b 1132,1133, 3

n actually presntt. The thydoid dec.........n factoea used fo this calcualation a-hall be these Iietod in Table 111 of TID 14844, AEG, 1062, "Calculation of Distacoe Fac-tors far Power and Test Reacto Site "!"OSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132. 1-133, 1-134, and I-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

(continued)

Catawba Units I and 2 1.1-1 Amendment Nos.-4""-

Definitions 1.1 1.1 Definitions (continued)

I DOSE EQUIVALENT XE-133 E-- AN VErAQE rIrl iK-Tr-t'%

AT-rtKl Ck1iCQrV DOSE EQUIVALENT XE-133 shall be that concentration of Xe-1 33 (microcunes 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-1 31m, Xe-133m, Xe-133, Xe-135m, Xe-1 35, 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-1 33 shall be performed using effective dose conversion factors for air submersion listed in Table 111. 1 of EPA Federal Guidance Report No.12, 1993, Extemal Exposure to Radionuclides in Air, Water, and Soil."

h h average (weighted in proportion to the

,%rna.~, ~

-Sr at ak

r. A;n

-164~

46 1-~n. 4aa

-4 the time of sampling) of the sum of the a...ag^ beta and gamma onorgies per dicintogration (in MSVMd) (Fr isotepes, o-ther than iodine, Wth half VlIve

-%10I

minutl, making up at loast 95% of the total ncniodi" e, a-;iviit in the coolant.

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE I.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued)

Catawba Units I and 2 1.1-2 Amendment Nos.44*t'ji-

RCS' Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The epec.fic

ti""ity of the.....

F

,..,.nt. Shall be MONhiA limits RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

APPLICABILITY:

MODES 1-and 2, 3, and 4.

I NIM-.-WU t 1

)

.t 8-.-.F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT

-Note-1-131 not within LCO 3.0.4.c is applicable.

limit..-* 4.0 "i/Cgm,.

A.1 Verify DOSE EQUIVALENT Once per4 hours 1-131 within the aGc.ptable ro.i"n of Figure 3.4. 16 1.

60 pCilgm.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

B.

GFc po.... a"ti';*

of Note-thO reacOtr coaPnt nRt LCO 3.0.4.c is applicable.

wthnA imrit. DOSE EQUIVALENT XE-133 not within limit.

B.1 Be in MODE 3 "with 648 hours0.0075 days 0.18 hours 0.00107 weeks 2.46564e-4 months T-5:4O9G-Restore DOSE EQUIVALENT XE-133 to within limit.

(continued)

I I

Catawba Units I and 2 3.4.16-1 Amendment Nos.--49f2e7-"

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion

-1,,- "00-r.

Time of Condition A Aet or B not met.

AND OR C.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT 1-131 > 60 pCi/gm. iR-the unaecoptablo rogion oa SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify...

tGo..

o.a. t grac, pe.ifir. at,,ty, 1 0l/6 In accordance with the Surveillance

-Note.......

Frequency Control Only required to be performed in MODE 1, 2, and 3 with Program RCS average temperature a 500 F.

Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity < 280 pCi/gm.

(continued)

Catawba Units 1 and 2 3.4.16-2 Amendment Nos 26"3,O

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.2 NO=rTE-f~.I.. ~

4~k

~

If A..

&Av MM 4 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm.

In accordance with the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period i

RR 3.1416.3 NOTE-Not rguFed to be perfomed u ntil3 ac fo miimu GOf 2 offctr.-G-i full poW8r days and 20 days oe ODE 1 Gopeation have elapsed stence the reactoF Was be~t rcu-boritical Ifo :1 8 houre.

Dete~e-~ fom a sample taken an MODE 1 afte-r-a m

inmm of 2 effoctivo full poWer days and 20 days 1MOGDE 1 oporatio haveo elapsed BMnW the reactor Was bect kqcrta fo 'a 18R ha, rs.

In Ramcrdanca Vith the SurvoillancoP Frequency ConrolG P, GO.

aM

~

u Catawba Units 1 and 2 3.4.16-3 Amendment Nos.-26M59-

PERCENT OF RATED THERMAL POWER Figure 3.4.16-1 (page 1 of 1)

Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER Catawba Units 1 and 2 3.4.16-4 Amendment Nos.-417-Y!6--

Definitions 1.1 1.1 Definitions I

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

,,8e DOSE EQUIVALENT 131 hal he that be etRatonoenf 1131 (microcueprgram) that alone would podupr thoe same thyroid does ai the quactity and iotipict mibtue of 11d31, 1

1 32,1 133,1

-34, and 11135 actually pracont. The thyroid dose carervemon factorce useod far. this-calcul-mdti-en shall be these fisted in Table III of T-1D 11844, AEG, 1062, "Calcbtionof1D1t1-8 FaGtore fo Power and Tot Reactor gke&.

7!!DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132,1-133, 1-134, and 1-135 actually present.

The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

I I

(continued)

McGuire Units 1 and 2 1.1-2 Amendment Nos. -.- /.-4e-

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT XE-1 33 6 -AVERA~GE DISINTEGRATION ENERGY DOSE EQUIVALENT XE-1 33 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 I11.1 of EPA Federal Guidance Report No.12,-93, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

-~

eh~

-seRGe I be the -aver-age (weighted in propertion to the ntration; of each FadionucWide In the reactor GOO laRt a 11 iI D 11 m

i I

mo rime or ~amniina or me ~um or me i~.roriae nemi inn gamma8 nerSgieG por dicintogration (in MoV'd) for icotOPpe, other than iod1noc, With half lioeS - 10 minuies, making up at leAmt OEM of thetoa nonio.io a*

7.m in t,*

coolant*.

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and (continued)

McGuire Units I and 2 1.1-3 Amendment Nos.-89-i RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The.p..ific.. *',ivity of the

,ea.t..

c.Gont Sl eh be MAthin I,.

RCS DOSE EQUIVALENT 1-131 and DOSE-EQUIVALENT XE-1 33 specific activty shall be within limits.

APPLICABILITY:

MODES-I, and-2, 3, and 4.

&A re 45

.& L b~1 I

AM*A I?

uj.

VA.0 MWORES93A TFIMARral.,

v-*

r v.

" -U ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT

-Note 1-131 not within limit.

LCO 3.0.4.c is applicable.

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131 withiR the 3cz~ptable region o Figuws 3.4.4-4.:

< 60pCi/gm.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

B.

GrFcO epG"oi.c,,ty of Note 6-48 hours the ro"cor G=,-.n. n*t LCO1 3.0.4.c is applicable.

wihin limit. DOSE EQUIVALENT XE-133 not within limit.

B.1 B in MODE 3 Wit-h T1-O44.04L Restore DOSE EQUIVALENT XE-133 to within limit.

(continued)

I I

McGuire Units 1 and 2 3.4.16-1 Amendment Nos. 224,.03 1

ACTIONS (continued)

RCS Specific Activity 3.4.16 CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3. with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B not met.

OR C.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT 1-131 iR-the unaccptablo rogion o

>igmwe4.164>

60 pJCilgm.

McGuire Units I and 2 3.4.16-2 Amendment Nos. 484466 I

RCS Specific Activity 3.4.16 Surveillance Requirements (continued)

SURVEILLANCE REQUIREMENTS I

SURVEILLANCE FREQUENCY 4.

SR 3.4.16.1

':crir: roaotor cooi~m aror~ wecaTic ~i~ir:

I r-j NOTE Only required to be performed in MODE 1 and 2, and 3 with RCS average temperature > 5001F.

In accordance with the Surveillance Frequency Control Program.

Verify reactor coolant DOSE EQUIVALENT XE-1 33 specific activity S 280 pCi/gm.

4.-

SR 3.4.16.2 wJniy Fegulr oa beQ peffenorma in ryinuD16 ;i.

In accordance with the Surveillance Frequency Control Program.

AND Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity 5 1.0 pCi/gm.

Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

I (continued)

McGuire Units 1 and 2 3.4.16-3 Amendment Nos. 484/4166

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY S-R 32.4.16.3 INOTE6 Not required to be potformad until 31 days 3fter a mi nm 2 e4ecti'o full powor days and 20 dayo ODE -

opefration have elapcsd since the reoacto was last subGritical for~l 48 hourc.

9eteaniRe

&fro a samnplo-takaen in 11.40E 1 aftor a m'inimum,,

of 2 offotwif full power day And 20 days of RA rr" 4

£

!A accordanAr vith the 2ur':OWlAWc Froqunocy Control' Peg@aM I

. ~

OR Affi-ItibmiGal W

-- me FIGUFS.

McGuire Units 1 and 2 3.4.16-4 Amendment Nos. 484466 I

RCS 250 o 200

'Ut.

ZE z LS~

Ii Y.

100 a

'U

~0 0a 50 0

20 30 40 50 60 70 80 PERCENT OF RATED THERMAL POWER Figuro 314169 1 (page 4 of 1)

Ra.-t," Co",,lat DOSE EQUIVALENT 11314 Spoifi Ati-l,,

I :=:&

ll------,---*

£

~A r

'1lEr*"t~ IillJiI~~l Xi*~

90

-IMil--A ~UAtt rcrMIuru at. F%%-v ~L

~~2:lawrAI&r McGuire Units I and 2 3.4.16-5 Amendment Nos. 1806t.6.

Definitions 1.1 1.1 Definitions CHANNEL CALIBRAT1ON (continued)

CHANNEL CHECK CHANNEL FUNCTIONAL TEST CONTROL RODS CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarms, interlocks, display, and trip functions.

CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1131 shall be that concenntioainfn of 1 r

31 (thiaturalones/am) that alno would produce !he iaeas thyreid dcmb -as the titieso and isotopic mixues of 1-31, I132, 1 133, 1134, and 1-135 actually present. The thyrid dotonsoe f

factor us-md far. th-a nalcubionA PhAll be these listed in Table Il of TI-D 146144, AEG, 19652, "CGalculation of Distance9 Factors fot Pamr arnd Tart Reantar Site;." DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from DOSE EQUIVALENT 1-131 OCONEE UNITS 1, 2. & 3 1.1-2 Amendment Nos. 3C,, 3S,, & 3

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE-133 Table 2.1 of the Environmental Protection Agency (EPA)

Federal Guidance Report No. 11.

DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (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 111. 1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil.'

Cr A'Vt-4ACL

.9a~Dotl avorario "~Il1mod in proporti ran,*I S,rr"/'nn~ S *=fl'.a.J I

p.

,rnfl.,

coolant at tho time of sampmig) of the sum of thon average beta and g m.

9nrgo per dicintgration (in Mo9169 for 0 F0,a making up at least 05% of the total noniodino activty in the eeolanM.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

OCONEE UNITS 1, 2, & 3 1.1-3 Amendment Nos. 35,,,,7, &,,,

RCS Specific Activity 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 RCS Specific Activity I

LCO 3.4.11 The speciAfc "c"i-ity of the...

ctor

.O..3nt shall b9 ""thin limits. RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPUCABILITY:

MODES 1-aAd 2, 3, and 4.

flh#'.r%

r%

ýPýt I

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE NOTE-EQUIVALENT 1-131 not LCO 3.0.4 is not applicable.

within limit.

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131

it.hn the a,,eptablo 5 50 pCi/gm.

AND 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months A.2 Restore DOSE EQUIVALENT 1-131 to within limit.

B.

Required Ati*on and-NOTE 42-.W..

a...

.,td Completion LCO 3.0.4 is not applicable.

Time of Conndition A not met-.DOSE EQUIVALENT XE-133 not within limit.

OR B.1

-Re iR MODE 3 With 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months RG.,veFsge DO-SE EQUIVALENT tempeFature 1 131 in unaccoptablo

'5000F.Restore DOSE OCONEE UNITS 1, 2, & 3 3.4.11-1 Amendment Nos. 300, 300, & 300

RCS Specific Activity 3.4.11 CONDITION REQUIRED ACTION COMPLETION TIME 3.4*4,4.

EQUIVALENT XE-1 33 to within limit.

(continued)

OCONEE UNITS 1, 2, & 3 3.4.11-2 Amendment Nos. 300, 300, &-30

RCS Specific Activity 3.4.11 CONDITION REQUIRED ACTION COMPLETION TIME C.

GFr-s pe.ific a*ivc

"*i C.1 Be in MODE 3. with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of tho coob t

nctwihin RGS Average, limit.Required Action

-mpr-twe e.o.

5002F.

and associated Completion Time of Condition A or B not AND met.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.2 Be in MODE 5 O__R DOSE EQUIVALENT I-131 > 50 pCi/gm SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1

\\'.,-f r.a. tor

,.,l...

gra,-

ep ifi. a;

ivi, I

NOTE.

_In accordance with the

-NOTE Only required to be performed in MODE 1, 2, Surveillance Frequency and 3 with RCS average temperature >500F.

Control Program Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity S 280 pCi/gm.

SR 3.4.11.2 Only raquired to be perftrmed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity

1.0 pCifgm.

In accordance with the Surveillance Frequency Control Program AND.

32

,&7 Amendment Nos. 372, 374., &_37 OCONEE UNITS 1, 2, & 3 3.4.11-3

RCS Specific Activity 3.4.11 SURVEILLANCE FREQUENCY Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period SR 3.4.11.3 NTE Not roquired to be porFormo~d until 31 dvAt

.r a minimu im of 2 EDr-,

and 20 day, "

MODE ! oporatiln have elapsed

'ino the roactor Wac be6-t 61ubcGAtritwal for Ž! 48 houre.

Determno

.~In arnordlanco w&ithth GunseillAnce rFrquonc" GOntrol Programnl84 days Amendment Nos. 372, 371, &

I OCONEE UNITS 1. 2. & 3 3.4.11-4

RCS a

10 0

0 lob Figure 3.4.11-1 (page 1 o Reactor C ant DOSE EQUIVALENT 1-131 S cific Activity Limit e*E THEMA Versus Pent of RATED THERMAL POWER Reactor Coolant S

ific Activity> 1.0 pClIgm DOSE EQUIV NT 1-131 OCONEE UNITS 1, 2, & 3 3.4.11-5 Amendment Nos. 328, 328 & 329 1 CNS, MNS, ONS Technical Specification and Bases Final Pages

Definitions 1.1 1.1 Definitions (continued)

I CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT XE-1 33 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

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 I

(continued)

Catawba Units 1 and 2 1.1-1 Amendment Nos.

471

Definitions 1.1 1.1 Definitions (continued)

I ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME EQUIVALENT XE-1 33 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No.12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System;

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; I

(continued)

Catawba Units 1 and 2 1.1-2 Amendment Nos. 17W17-1

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

f t

APPLICABILITY:

MODES 1 2,3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT Note 1-131 not within limit.

LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1-131 < 60 pCVgm.

AND A.2 Restore DOSE EQUIVALENT 1-131 to within limit.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B.

DOSE EQUIVALENT Note XE-133 not within limit.

LCO 3.0.4.c is applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-133 to within limit.

(continued)

Catawba Units I and 2 3.4.16-1 Amendment Nos. 243!207 I

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or AND B not met.

C.2 Be in MODE 5.

OR 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT 1-131 > 60 pCilgm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Note-In accordance with Only required to be performed in MODE 1, 2, and 3 with the Surveillance RCS average temperature > 500 F.

Frequency Control Program Verify reactor coolant DOSE EQUIVALENT XE-1 33 specific activity ;280 pCi/gm.

(continued)

I Catawba Units 1 and 2 3.4.16-2 Amendment Nos. 263Q59 I

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

I SURVEILLANCE FREQUENCY SR 3.4.16.2 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm.

In accordance with the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period Catawba Units 1 and 2 3.4.16-3 Amendment Nos.-26@OK*6

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT XE-1 33 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A COT shall be the injection of a simulated or actual signal into the channel as dose to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

DOSE EQUIVALENT XE-133 shall be that concentration of Xe-1 33 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nudides 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 111.1 of EPA Federal Guidance Report No. 12, 1993, 'External Exposure to Radionuclides in Air, Water, and Soil.

(contnued (continued)

McGuire Units I and 2 1.1-2 Amendment Nos.

RCS Specific Activily 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

I J

APPLICABILITY:

MODES 1,2,3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT No%

1-131 not within limit.

LCO 3.0.4.c is applicable.

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131

<60pCi/gm.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

B.

DOSE EQUIVALENT Note XE-133 not within limit.

LCO 3.0.4.c is applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-1 33 to within limit.

(continued) i.

McGuire Units I and 2 3.4.16-1 Amendment Nos.

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be In MODE 5 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR DOSE EQUIVALENT 1-131 >60pCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 NOTE Only required to be performed in MODES 1, 2, and 3 with In accordance RCS average temperature > 500°F.

with the Surveillance Frequency Control Verify reactor coolant DOSE EQUIVALENT XE-133 Program specific actity< 280 pC/gm.

SR 3.4.16.2 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity <1.0 pCi/gm.

In accordance with the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of_> 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period

-a I

McGuire Units I and 2 3.4.16-2 Amendment Nos.

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CALIBRATION (continued)

CHANNEL CHECK CHANNEL FUNCTIONAL TEST CONTROL RODS CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarms, interlocks, display, and trip functions.

CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

OCONEE UNITS 1, 2, & 3 1.1-2 Amendment Nos. 355, 35:7, &

I

Definitions 1.1 1.1 Definitions (continued)

CONTROL RODS CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT XE-133 CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131,1-132,1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

DOSE EQUIVALENT XE-133 shall be that concentration of Xe-1 33 (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 I11.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

OCONEE UNITS 1, 2, & 3 1.1-3 Amendment Nos. 366, 368 & 367

RCS Specific Activity 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 RCS Specific Activity LCO 3.4.11 The RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

APPLICABILITY:

MODES 1 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE NOTE EQUIVALENT 1-131 not LCO 3.0.4 is not applicable.

within limit.

A.1 Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131

5 50 pCi/gm.

AND A.2 Restore DOSE EQUIVALENT 1-131 to within limit.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B.

DOSE EQUIVALENT NOTE XE-133 not within limit.

LCO 3.0.4 is not applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-133 to within limit.

(continued)

II OCONEE UNITS 1, 2, & 3 3.4.11-1 Amendment Nos. 300, 300, & 300

RCS Specific Activity 3.4.11 CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met.

AND OR C.2 Be in MODE 5 DOSE EQUIVALENT I-36 hours 131 > 50 pCi/gm SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 NOTE Only required to be performed in MODE 1, 2, and 3 with RCS average temperature ;500F.

In accordance with the Surveillance Frequency Verify reactor coolant DOSE EQUIVALENT Control Program XE-133 specific activity s 280 pCi/gm.

SR 3.4.11.2 Verify reactor coolant DOSE EQUIVALENT In accordance with the 1-131 specific activity < 1.0 l+/-Ci/gm.

Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after THERMAL POWER change of >_ 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period I

OCONEE UNITS 1, 2, & 3 3.4.11-2 Amendment Nos. 37-2, 3,7, & 37 I

SComplete Replacement of the Existing RCS Specific Activity B 3.4.16 3.4.16 Bases B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1) or 10 CFR 100.11 (Ref. 5). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria.

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite and control room doses meet the appropriate SRP acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gpd per SG exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 ACi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.17, "Secondary Specific Activity."

The analyses for the SLB and SGTR accidents establish the acceptance limits for RCS specific activity. Reference to these analyses is used to Catawba Units I and 2 B 3.4.16-1 Revision No. 2

RCS Specific Activity B3.4.11 BASES (continued) assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 JCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60 iLCi/gm DOSE EQUIVALENT 1-131 due to an iodine spike caused by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 280 pCi/gm DOSE EQUIVALENT XE-1 33.

The SGTR analysis also assumes a loss of offsite power at the same time as the reactor trip. The SGTR causes a reduction in reactor coolant inventory.

The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves.

The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the Residual Heat Removal (RHR) system is placed in service.

The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the RHR system is placed in service.

Operation with iodine specific activity levels greater than the LCO limit is permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if the activity levels do not exceed 60.0 pCi/gm.

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The iodine specific activity in the reactor coolant is limited to 1.0 PCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the Catawba Units I and 2 B 3.4.16-2 Revision No.

I

RCS Specific Activity B3.4.11 BASES (continued) reactor coolant is limited to 280 pCi/gm DOSE EQUIVALENT XE-133.

The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

The SLB and SGTR accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).

APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 is necessary to limit the potential consequences of a SLB or SGTR to within the SRP acceptance criteria (Ref. 2).

In MODES 5 and 6, the steam generators are not being 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.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the specific activity is < 60.0 pCi/gm. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Actions A.1 and A.2 while the DOSE EQUIVALENT 1-131 LCO limit is not met.

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 Catawba Units I and 2 B 3.4.16-3 Revision No.

I

RCS Specific Activity B3.4.11 BASES (continued) proceeds to, power operation.

B.1 With the DOSE EQUIVALENT XE-1 33 greater than the LCO limit, DOSE EQUIVALENT XE-1 33 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-1 33 LCO limit is not met. 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.

C.1 and C.2 If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is > 60.0 pCi/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Catawba Units I and 2 B 3.4.16-4 Revision No.

I

RCS Specific Activity B3.4.11 BASES (continued) ff a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

A Note modifies the SR to allow entry into and operation in MODE 4, MODE 3 with RCS average temperature < 500F prior to performing the SR. This allows the establishment of the necessary plant conditions to produce sufficient sample flow.

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change >

15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

REFERENCES

1. 10 CFR 50.67.

2.

Standard Review Plan (SRP) Section 15.0.1 "Radiological Consequence Analyses Using Alternative Source Terms."

3. FSAR, Section 15.1.5.

4. FSAR, Section 15.6.3.

5.

10CFR 100.11.

Catawba Units I and 2 B 3.4.16-5 Revision No.

I

RCS Specific Activity Complete Replacement of the Existing BR3.4.16 3.4.16 Bases B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 100.11 (Ref 4) or 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria.

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite and control room doses meet the appropriate acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 2 and 3) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant total steam generator (SG) tube leakage rate of 389 gpd exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 AiCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.16, "Secondary Specific Activity."

The analyses for the SLB and SGTR accidents establish the acceptance McGuire Units 1 and 2 B 3.4.16-1 Revision No-mff-

RCS Specific Activity B.3.4.16 BASES (continued) limits for RCS specific activity. Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.01iCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60 jLCi/gm DOSE EQUIVALENT 1-131 due to an iodine spike caused by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 280pCi/gm DOSE EQUIVALENT XE-133.

The SGTR analysis also assumes a loss of offsite power at the same time as the reactor trip. The SGTR causes a reduction in reactor coolant inventory.

The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves.

The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the Residual Heat Removal (RHR) system is placed in service.

The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the RHR system is placed in service.

Operation with iodine specific activity levels greater than the LCO limit is permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if the activity levels do not exceed 60.0 pCi/gm.

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

McGuire Units I and 2 B 3.4.16-2 Revision No. -ff-

RCS Specific Activity B.3.4.16 BASES (continued)

LCO The iodine specific activity in the reactor coolant is limited to 1.0 ipCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 280 pCi/gm DOSE EQUIVALENT XE-1 33.

The limits on specific activity ensure that offsite and control room doses will meet the appropriate acceptance criteria.

The SLB and SGTR accident analyses (Refs. 2 and 3) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the acceptance criteria.

APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 is necessary to limit the potential consequences of a SLB or SGTR to within the acceptance criteria.

In MODES 5 and 6, the steam generators are not being 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.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the specific activity is < 60.0 pCi/gm. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Actions A.I and A.2 while the DOSE EQUIVALENT 1-131 LCO limit is not met.

McGuire Units I and 2 B 3.4.16-3 Revision No-tf-

RCS Specific Activity B.3.4.16 BASES (continued)

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.

B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-1 33 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there was a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Action B. 1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. 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.

C.1 and C.2 If the Required Action and associated Completion lime of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is > 60 pCilgm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

McGuire Units I and 2 B 3.4.16-4 Revision No.-&76,

RCS Specific Activity B.3.4.16 BASES (continued)

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-1 33 is not detected, it should be assumed to be present at the minimum detectable activity.

A Note modifies the SR to alloW entry into and operation in MODE 4, MODE 3 with RCS average temperature < 500F prior to performing the SR. This allows the establishment of the necessary plant conditions to produce sufficient sample flow.

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change >

15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

REFERENCES

1.

10CFR50.67.

2.

FSAR, Section15.1.5

3. FSAR, Section 15.6.3.

4. 10CFR100.11 McGuire Units I and 2 B 3.4.16-5 Revision No.aff-

Complete Replacement of the Existing 3.4.11 Bases B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 RCS Specific Activity BASES RCS Specific Activity B 3.4.11 BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 100.11 (Ref 1) or 10 CFR 50.67 (Ref.4). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33. The allowable levels are intended to ensure that doses meet the appropriate acceptance criteria.

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed the 10 CFR 100.11 (Ref. 1) or 10 CFR 50.67 (Ref. 4) dose guideline limits following an SGTR or a steam line break (SLB) accident. The SLB safety analysis (Ref. 2) assumptions bound the specific activity of the reactor coolant at the LCO limits and a total existing reactor coolant steam generator (SG) tube leakage rate of 300 gpd.

The analysis results are significantly impacted by the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the facility that could affect RCS specific activity as they relate to the acceptance limits.

I OCONEE UNITS 1, 2, & 3 B 3A.11-1 Amendment Nos..31O, 3O,- &4

RCS Specific Activity B 3.4.11 I BASES (continued)

APPUCABLE The safety analysis shows the radiological consequences of an SLB and SAFETY ANALYES SGTR accident are within the dose guideline limits. Operation with iodine (continued) specific activity levels greater than the LCO limit is permissible, If the activity levels do not exceed the limits for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

RCS Specific Activity satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3).

LCO The specific iodine activity is limited to 1.0 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity In the primary coolant is limited to 280 iCi/gm DOSE EQUIVALENT XE-1 33. The limits on specific activity ensure that the doses will meet the appropriate acceptance criteria.

The SGTR and SLB accident analyses (Ref. 2) shows that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR or SLB, lead to site boundary doses that exceed the NRC dose guideline limits.

APPLICABILITY In MODES I and 2, 3 and MODE 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 are necessary to limit the potential consequences of an SGTR or SLB to within the acceptable site boundary dose values.

In MODES 5 and 6, the steam generators are not being 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..

I OCONEE UNITS 1, 2, & 3 B 3.4.11-2 Amendment Nos. 3800,030, & 3

RCS Specific Activity B 3.4.11 BASES (continued)

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the specific activity is < 50pCI/gm.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Is required to obtain and analyze a sample. Sampling must continue for trending.

The DOSE EQUIVALENT 1-131 must be restored to limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

A Note to RA A.1 and A.2 excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODES(s) while relying on the ACTIONS even though the ACTIONS may eventually require unit shutdown. This exception 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 unit remains at, or proceeds to power operation.

B.1I With the DOSE EQUIVALENT XE-133 greater than the LCO limit. DOSE EQUIVALENT XE-1 33 must be restored to within imit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note to RA B.1 excludes the MODE change restriction of LCO 3.0.4.

This exception allows entry into the applicable MODES(s) while relying on the ACTIONS even though the ACTIONS may eventually require unit shutdown. This exception 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 unit remains at, or proceeds to power operation.

C.1 and C.2 If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is > 5OpCi/gm, the reactor must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating OCONEE UNITS 1, 2, & 3 B 3.4.11-3 Amendment Nos.

I

RCS Specific Activity B 3.4.11 BASES (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS SR 3.4.11.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This surveillance provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-1 33 is not detected, it should be assumed to be present at the minimum detectable activity.

A note modifies the SR to allow entry into and operation in MODE 4, MODE 3 with RCS average temperature < 500F prior to performing the SR. This allows the establishment of the necessary plant conditions to produce sufficient sample flow.

SR 3.4.11.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change a15% RTP within a I hour period is established because the iodine levels peak during this time foIlowing iodine spike initiation; samples at other times would provide inaccurate results.

OCONEE UNITS 1,2, &3 B.3.4.11-4 Amendments Nos. COO. 3C0, &S

RC-S Specific Activity B 3.4.11 BASES (continued)

REFERENCES

1.

10 CFR 100.11.

2.

UFSAR, Section 15.9, 15.13, and 15.17

3.

10 CFR 50.36.

4.

10 CFR 50.67 II I

OCONEE UNITS 1, 2, & 3 B 3.4.11-5 Amendment Nos. BIB, 800, &-See

ML11271A027

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ML11271A027

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