1CAN030802, License Amendment Request Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec

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License Amendment Request Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec
ML080850906
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/13/2008
From: Mitchell T
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN030802, TSTF-490, Rev 0
Download: ML080850906 (28)
v • d • e


Text

4 Entergy Operations, Inc.

-E)7 1448 SýR. 333 Russellville, AR 72802 Tel 479-858-3110 Timothy G. Mitchell Vice President, Operations Arkansas Nudclear One 1CAN030802 March 13, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" Arkansas Nuclear One, Unit 1 Docket Nos. 50-313 License Nos. DPR-51

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment to the Technical Specifications (TS) for Arkansas Nuclear One, Unit 1.(ANO-1).

The proposed changes would replace the current pressurized water reactor (PWR) TS 3.4.12 (ANO-1) limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-1 33 (DEX) definition that would replace the current E Bar average disintegration energy definition. In addition, the current dose equivalent 1-131 (DEI) definition would be revised to allow the use of additional thyroid dose conversion factors (DCFs).

The changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." The availability of this TS improvement was announced in the Federal Register on March 15, 2007 (72FR12217) as part of the consolidated line item improvement process (CLIIP). provides a description and assessment of the proposed changes, as well as confirmation of applicability. Attachment 2 provides the existing TS pages marked-up to show the proposed changes. Attachment 3 provides final (clean) TS pages. Attachment 4 provides a markup of the associated TS Bases, for information only.

The proposed changes do not include any new commitments.

4~ci

f 1CAN030802 Page 2 of 2 Entergy requests approval of the proposed amendment by October 1, 2008. Once approved, the amendment shall be implemented within 90 days. Although this request is neither exigent nor emergency, your prompt review is requested. In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Arkansas State Official.

If you have any questions or require additional information, please contact David Bice at 479-858-5338.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 13, 2008.

Sincerely, TGM/dbb Attachments:

1. Description and Assessment of Proposed Changes
2. Proposed Technical Specification Changes
3. Final Technical Specification Pages
4. Technical Specification Bases Changes Markups (For Information Only) cc:
  • Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS 0-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health & Human Services P.O. Box 1437 Slot H-30 Little Rock, AR 72203-1437

Attachment I ICAN030802 Description and Assessment of Proposed Changes

Attachment to 1CAN030802 Page 1 of 4

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-51 for Arkansas Nuclear One, Unit 1 (ANO-1).

The proposed changes would replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-1 33 and would take into account only the noble gas activity in the primary coolant. The changes were approved by the NRC staff Safety Evaluation (SE) dated September 27, 2006 (ADAMS ML062700612) (Reference 1). Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 15, 2007 as part of the consolidated line item improvement process (CLIIP).

2.0 PROPOSED CHANGE

Consistent with NRC-approved TSTF-490, Revision 0, the proposed TS changes:

Revise the definition of DOSE EQUIVALENT 1-131.

Delete the definition of E-AVERAGE DISINTEGRATION ENERGY.

Add a new TS definition for DOSE EQUIVALENT XE-1 33.

Revise LCO 3.4.12, "RCS Specific Activity" to delete references to gross specific activity and add limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33. (Note that ANO-1 TSs do not contain a DOSE EQUIVALENT 1-131 versus percent power figure noted for deletion in TSTF 490)

Revise LCO 3.4.12 "Applicability" to specify the LCO is applicable in MODES 1, 2, 3, and 4.

Modify ACTIONS Table as follows:

A.

Condition A is modified to replace "specific activity" with "DOSE EQUIVALENT 1-131" and define an upper limit for DOSE EQUIVALENT 1-131 that is applicable at all power levels.

B. ACTIONS are reordered, moving Condition B to Condition C to be consistent with the Writer's Guide for the improved TSs of NUREG 1430.

C.

A new Condition B is added to provide a Condition and Required Action for DOSE EQUIVALENT XE-133 with a Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A Note allowing the applicability of LCO 3.0.4.c is added, consistent with the Note to Required Action A.1 described above.

D. Condition C (was Condition B) is modified based on the changes to Conditions A and B and to reflect the change in the LCO Applicability.

Revise SR 3.4.12.1 to verify the limit for DOSE EQUIVALENT XE-133. A Note is added to allow entry into MODES 2, 3, and 4 prior to performance of the SR.

" Delete SR 3.4.12.3.

J

Attachment to 1CAN030802 Page 2 of 4 provides a markup of the affected TSs delineated above. Attachment 3 provides the final (clean) TS pages associated with these changes. includes a markup of the impacted TS Bases pages for information only. The TS Bases will be revised in accordance with the TS Bases Control Program (TS 5.5.14) during implementation of TSTF-490 once approved by the NRC. Also note that the TS Bases markup captures the information necessary to support the ANO-1 Alternate Source Term (AST) submittal discussed below. Therefore, the wording deviates from that proposed in TSTF-490, but is consistent with the Licensing Basis assuming approval of AST for ANO-1.

Note that TSTF-490 was written and approved considering the most current version of the Improved Technical Specifications (ITS) with approved TSTFs at the time. As described above, this included reference to Limiting Condition for Operation (LCO) 3.0.4.c, which was incorporated in the ITS under TSTF-359, Mode Change Limitations. Entergy has submitted a proposal to adopt TSTF-359, which is currently under review by the NRC (Reference 4).

Therefore, the above changes associated with LCO 3.0.4 are dependent on the NRC approval of Entergy's TSTF-359 adoption proposal.

In addition to the discussion of TSTF-359 above, Entergy has also submitted an application to adopt use of Alternate Source Term (AST) for ANO-1 (Reference 5). This application is currently under NRC review and affects the pages associated with TS 3.4.12 above. In support of both TSTF-490 and AST adoption, the DOSE EQUIVALENT 1-131 Definition is modified to refer to the committed effective dose equivalent (CEDE), which is consistent with AST.

Therefore, the markup and final versions of TS 3.4.12 included in Attachments 2 and 3, and the TS 3.4.12 Bases markup included in Attachment 4 are shown as if the ANO-1 TSTF-359 and the AST applications have been approved (minus the addition of a new amendment number in the associated page footer). Entergy will submit new markup and final TS pages upon NRC request should either the TSTF-359 or AST application be revised or should the approval of either of these applications be significantly delayed. 'If the aforementioned applications are approved as anticipated, Entergy will submit new final TS pages that include the new amendment numbers upon request from the NRC.

3.0 BACKGROUND

The background for this application is as stated in the model safety evaluation (SE) in the NRC's Notice of Availability published on March 15, 2007 (72FR12217), the NRC Notice for Comment published on November 20, 2006 (71 FR67170), and TSTF-490, Revision 0.

4.0 TECHNICAL ANALYSIS

In the model SE, the NRC included statements which would require the licensee to identify specific information in support of adopting TSTF-490. The following provides ANO-1 specific information in this regard.

1.

Section 3.1.1 of the model SE includes a list of acceptable dose conversion factors (DCF) for use in the determination of dose equivalent iodine (DEI) in relation to dose consequence analyses. The ANO-1 analyses employ Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

Attachment to 1CAN030802 Page 3 of 4

2.

In the first paragraph of Section 3.1.2 of the model SE, a bracketed list of isotopes is provided that designates the noble gases that may be used in the determination of dose equivalent xenon (DEX). All isotopes depicted within these brackets are currently considered in the ANO-1 calculation of DEX. This is captured in the proposed ANO-1 definition of DEX to be included in TS Section 1.1.

3.

Section 3.1.2 of the model SE also provides two possible determination methods for DEX.

ANO-1 uses the effective dose conversion factor 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." This is also depicted in the proposed definition of DEX in TS Section 1.1.

4.

Section 3.1.3 states that it is incumbent on the licensee to ensure that the site specific limits for both DEI and DEX are consistent with the current steam generator tube rupture (SGTR) and main steam line break (MSLB) radiological consequence analyses. The ANO-1 analyses value for DEI is 1.0 pCi/gm as depicted in the proposed TS markup and also provided to the NRC in the aforementioned AST submittal (Reference 5). The ANO-1 analyses value for DEX is 2200 pCi/gm, considering the adoption and use of AST discussed previously. This is also depicted in the attached TS markups.

As noted in the Section 3.0 above, the DOSE EQUIVALENT 1-131 Definition depicted in TSTF-490 is modified to refer to the committed effective dose equivalent (CEDE), which is consistent with AST. This change will remove any conflict between the adoption of TSTF-490 and the adoption of AST for ANO-1. Entergy does not consider this a significant deviation from the TSTF-490 adoption under CLIIP.

Entergy Operations, Inc. (Entergy) has reviewed References 1, 2 and 3, and the model SE published on November 20, 2006 (71 FR67170) as part of the CLIIP Notice for Comment.

Entergy has applied the methodology in Reference 1 to develop the proposed TS changes.

Entergy has also concluded that the justifications presented in TSTF-490, Revision 0 and the model SE prepared by the NRC staff are applicable to ANO-1 and justify this amendment for the incorporation of the changes to the ANO-1 TS.

5.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on March 15, 2007 (72FR12217), the NRC Notice for Comment published on November 20, 2006 (71 FR67170), and TSTF-490, Revision 0.

6.0 NO SIGNIFICANT HAZARDS CONSIDERATION Entergy Operations, Inc. (Entergy) has reviewed the proposed no significant hazards consideration determination published in the Federal Register on March 15, 2007 (72FR12217) as part of the CLIIP. Entergy has concluded that the proposed determination presented in the notice is applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

Attachment to 1CAN030802 Page 4 of 4 7.0 ENVIRONMENTAL EVALUATION Entergy Operations, Inc. (Entergy) has reviewed the environmental consideration included in the model SE published in the Federal Registeron March 15, 2007 (72FR12217) as part of the CLIIP. Entergy has concluded that the staffs findings presented therein are applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and the determination is hereby incorporated by reference for this application.

8.0 REFERENCES

1.

NRC Safety Evaluation (SE) approving TSTF-490, Revision 0 dated September 27, 2006

2.

Federal Notice for Comment published on November 20, 2006 (71 FR 67170)

3.

Federal Notice of Availability published on March 15, 2007 (72FR12217)

4.

Entergy letter to NRC dated October 22, 2007, Technical Specification Changes Regarding, Mode Change Limitations and Associated Bases Using the Consolidated Line Item Improvement Process (TSTF-359), Arkansas Nuclear One, Unit 1 (1CAN100701)

(TAC No. MD7179)

5.

Entergy letter to NRC dated October 22, 2007, Technical Specification Changes and Analyses Relating to Use of Alternate Source Term, Arkansas Nuclear One, Unit 1 (1CAN100703) (TAC No. MD7177) 1 CAN030802 Proposed Technical Specification Changes

Definitions 1.1 1.1 Definition CHANNEL CALIBRATION (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.

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 of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps.

CONTROL RODS shall be all full length safety and regulating rods that are used to shutdown 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 ANO-1 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. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1 131 6hall be tha;t c-nc--entratio-n of I 131 (microcurFie6!gram) that alone would preduca8 the Game thyrofid dose as the quantity and isotopic mqfixture at 1131,1 132,1 133, 113, ad 135 atually p*roAnt. The thyroid dosGe oRnVersioR fa(ctors useld fo-r thi-calcu-latioR shall be those listed in Table !I! of TID 14844, AEC, 1962, 'CalculationR of Distancc F~actors for Power and Test Reactor Sites." DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per qram) that alone would produce the same committed effective dose eauivalent (CEDE) as the auantitv and isotopic mixture of DOSE EQUIVALENT 1-131 ANO-1 1.1-2 Amendment No. 2-1-5, ANO-1 1.1-2 Amendment No. 245,

Definitions 1.1 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The CEDE dose conversion factors used to determine the DOSE EQUIVALENT 1-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

ANO-1 1.1-2 Amendment No. 245, ANO-1 1.1-2 Amendment No. 245,

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT XE-1 33 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-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. 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, "External Exposure to Radionuclides in Air. Water, and Soil."

-AVE RAGEF L.';*.';%' : *L*.'*.."*.

  • .*_'.* *.*_."'.k*.

LEAKAGE Eshall be tho average (weighted in proportion to the concentration of cach radionuc~lide in the reactorF coolant at the time of sampnlnng) of tho sunm of the avorago beta and gamma ontrgs por 9FtdisitegrFation (in Mo]V) for io

,tpes,,

,t*hr,*,tha

iodne, with half IY,6,v 15 mOinuts, r;aking up at cast of the total no*nidine actiffity in the G004Pit.

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);
b.

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

c. Pressure Boundary LEAKAGE ANO-1 1.1-3 Amendment No. 2--5,224,

Definitions 1.1 LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 1with fuel in the reactor vessel.

MODE To be moved to next page ANO-1 1.1-3 Amendment No. 2-1-fr~.24, ANO-1 1.1-3 Amendment No. 245,2-24,

RCS Specific Activity 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 RCS Specific Activity LCO 3.4.12 TheRCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity of the reactor coolant shall be within limits.-

a.

5 "1.0 pi/gm DOSE EQUIVALENT 1 131; and

b. :- 724i pi/gm total.

APPLICABILITY:

MODES 1, aA4d-2, 3, and 4 Pr

\\ '~ rAAOI~

4.

VVt*

LT A

fllýft r

no ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT NOTE-------

1-131S not LCO 3.0.4.c is applicable.

within limits.

A.1 Verify DOSE EQUIVALENT Once per4 hours 1-131 -< 60 IJCi/qm.Reste)-e specific activity to Within AND A.2 Restore DOSE 248 hours0.00287 days <br />0.0689 hours <br />4.100529e-4 weeks <br />9.4364e-5 months <br /> EQUIVALENT 1-131 to within limit.

B.

DOSE EQUIVALENT NOTE XE-1 33 not within limit.

LCO 3.0.4.c is applicable.

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

ANO-1 3.4.12-1 Amendment No. 2-1-5,

RCS Specific Activity 3.4.12 C9. Required Action and C8.1 Be in MODE 3-with-:,

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion

<-5000F.

Time of Condition A or B not met.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1

NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 7 days XE-1 33ger-&

specific activity < 22007-2/R_ pCi/gm.

SR 3.4.12.2

NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity < 1.0 pCi/gm?

S R 3.4.!2.3 NOTE Not requir~ed to be performed until 31 days aftor a minimum of 2 EF=PD and 20 days ef MOIDE I op.ratiOn have elapsed since the..

actor was last UbcriFticl-for Ž 48 hou*rs Determne-F 7 8A4days ANO-1 3.4.12-2 Amendment No. 24-5, 1 CAN030802 Final Technical Specification Pages

Definitions 1.1 1.1 Definition 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.

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 of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps.

CONTROL RODS shall be all full length safety and regulating rods that are used to shutdown 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 ANO-1 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. 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 committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT 1-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

ANO-1 1.1-2 Amendment No. 2-1-5, ANO-1 1.1-2 Amendment No. 245,

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT XE-1 33 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-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."

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);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE;
c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

ANO-1 1.1-3 Amendment No. 245,224, ANO-1 1.1-3 Amendment No. 24-5,2-24,

I..

Definitions 1.1 1.1 Definition (continued)

MODE OPERABLE-OPERABILITY PHYSICS TESTS A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or.

device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a.

Described in the SAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

QPT shall be defined by the following equation and is expressed as a percentage.

QUADRANT POWER TILT (QPT)

OPT = 100 ( Power in any Core Quadrant QAverage Power in all Quadrants RTP shall be a total steady state reactor core heat transfer rate to the reactor coolant of 2568 MWt.

1)

RATED THERMAL POWER (RTP)

ANO-1 1.1-4 Amendment No. 2-1-5,224, ANO-1 1.1-4 Amendment No. 2-1-5,2-24,

RCS Specific Activity 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 RCS Specific Activity LCO 3.4.12 APPLICABILITY:

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

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 per4 hours 1-131 < 60 pCi/gm.

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

B.

DOSE EQUIVALENT

NOTE ---------

XE-1 33 not within limit.

LCO 3.0.4.c is applicable.

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

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

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

ANO-.1 3.4.12-1 Amendment No. 245,

RCS Specific Activity 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1

NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-i 33 7 days specific activity < 2200 pCi/gm.

SR 3.4.12.2

NOTE Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 14' days specific activity < 1.0 pCi/gm.

ANO-1 3.4.12-2 Amendment No. 24-5, 1CAN030802 Technical Specification Bases Changes Markups (For Information Only)

RCS Specific Activity B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM B 3.4.12 RCS Specific Activity BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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). 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 oAde

,,f Fodeal Regulations, 10 CPR 100 (Ref. 1), Gpoc3ifier, the ma.ximumP doso to tho Axholo body and the thyroid -an findiVidul-al at tho--99 sh boundary can r.ceiVe for 2 h.ur. dUring an accido;nt The limit on... pecific activity ensur. that the doses aro hold to a smallfc.tio.n of th.

.10 CFR 100 limits during analyzed transicnts and accidents.

The RCS specific activity LCO limits the allowable concentration ievel of radionucGlid8esi the reactor coolant. The LCO limits are established to minimize the offsitc radioactivity dose co9nsequences in the event of a steam generator tube rupture (SGT-R) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33.The C c,ontai*n, specific aGtivity limitS for b.th DOE EQUIVALENT 1131 and total specifi activity. The allowable levels arc irtended to limit the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the site boundary to a small fraction of the 10 CFR 100 dose APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite and control room doses will not exceed the applicable 10 CFR 50.67 requirements following a steam line break (SLB) or steam generator tube rupture (SGTR) accident. The safety analyses 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 steam generator exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 ICi/qm DOSE EQUIVALENT 1-131 from LCO 3.7.4, "Secondary Specific Activity."The LCO limits on the specific activity of the reactor coolant enrsure that the resulting 2 howr doses at the site bounday Will n.ot exceed a small fr*ati-;n of the 10 CFR 100 dose guideline limits fo)llo)Wing an SGT-R acident. The thyroid dose conv8ersionR factorsE used'in the calcu,,lation of DOSE EQUIVALENT 1 131 are identified in SectiEo 1.1, "Definitions."

The analyses for the SLB and SGTR accidents establish the acceptance 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.

ANO-1 B 3.4.12-1 Amendment No. 215 Rev.

RCS Specific Activity B 3.4.12 Rupture of a steam gene-rattor tubbe wiould allow primnar' coolant actiVity to cntor the seconary colah. Th m1 FajorF porion of thiG activity is noble gases and would be released to the atmosphere f&Ro the condenser vacu1um pump or a relief valve. Activity would continue to be released until the operator could reduce the primary system pressure below the setpgint o'f the secndary relief valves. and coul.d isolate the faulty steam generator. The wo-rs.t caredible_ set of c!Gircnmtances is consider-ed-to -he a douible ended break of a Single lteam g.eneato. r be, followed by isol.ati*o* Af the faulty steam geReFrator within 34 mhiutes aft& the tube break. Assumirg the full differential Pe*SSUre across thep team generaFto, no m*ee than on* quarter of the total primary co;lAnt could be released to the secndary eoolant in this period. The decay h dits od 1 ho fr pessre reS e

edUcton. Will generate steam in the Seondawrysystem representing less than 15 weight percent of the secondary system.

The parameters assumed inthe doso analysis (Ref. ) for the single steam generator tube failureG included thefolwnvaus 1

p.

total priryolant vrlum V

e (mass) - 5.2 x 1 lbs-

2.

total secondary coolant voluIme (mass)

-2 X1

-te

3.

leakage rate from primary to Se.ondary sytem - 1 gpm.

4.

fiGs6io

  • pro*dut decayheat energy for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> -

1.56 I 08-

5.

steam mass released-to envi~ron 2.8 w 1Q06-flqP

6.

primaJ

-E7Olavt released to secondarY (34 minutes) - 8.7 x 1044 1 f

7.

minimu1m p rimary to secondary ioie9eqilibrium; activity ratio - 20 te 1 (for 1 gpmR leakage)-.

8.

D.EQUIVALENT 1131 specific activity - 3.5 P.gm. (

9.

DOSE EQUIVALENT 1 131 specific activity - 0.17 P~i/gm (Secondary).

10.

total specific activity i primary - 721F= p~i/gm-.

41. l Q -7
1.

0

-4 GV G /m lat limiting ploit beyond sit I

oundary of 1046 meters for 3_0 M. release height equivalent to ground level release due to topography including building wake Geffet forF 5 percentile meteorology.

12.

total ra dioactivity in primnary coolant released to s8eondary coolant reloased to

13.

ten perc.nt o. f the o.bined radi..odin. activity from prim..ay P.t.vity in secndarY coolant and secondary activity present isteam mass (released to environs) assumed released to environs.

The whole body dose rosulting from im~eme)rso in the cloud containing the released activity would include both gamma and beta radiation. The garnmma do-se is, dependent on the-finite-sizoand configuration of the cloud. However, the analysis emplyed the simple model of the semi infinite clou d, which gives an upper limit to the potential B 3.4.12-2 A.MenRdmenA Rt N o. 2 15 Rev.

ANO-1

RCS Specific Activity B 3.4.12 gamma dose. The semi infinite cloud model is applicable to the beta dose, because of the shEod range of beta radiation in air. The resulting whole body dose was determined to be lss than 0.5 Rem for this accident.

The thyroid deoqe fromn the steam; generator tube rupture accide-nt has, beenanalyzed aumIng I*

Ftu*be ruptur at fulll lo*ad*

d loesrs f o,, ite power at the time of the rea*t*r trip, which result inR steam release through the ý elif valves in the peFiod before the faulty steamn generator is isolated and primary' system pressure is reduced. The lifmiting iodiRe actiVities for the primary and secondany systems are used RI the initial GcRditeoRn.

One tenth of the *Eodine conRtained in the liquid which mrs conveold to steamn and passed through the relief valves is assumed to reach the site boundary'. The resul1ting thyroi*d dose fromF the comFbined primary' and secondar,' iodine activity released to the enviFros was determnined to be 1.5 Rem forF this accidbnt.

The_ limiAt forF secondary iodine activity iscnssen ith tho limlits onprm ryytemR iodine activity and primary to secondary leakage of 1 gpmn. if thciitysould exceed the specified limits follownapoe transient, the major concern would be whether additional fuel defects had developed bringing the total to above ex(pected level.Fo the obsor.'od removal of excess activity by decay and cleanup, it should be apparent whether actfivity is returning to a level below the specfificationR limit. Appropriate action to be taken to bring the activity within specification include onRe Or More of the following:

gradual decreasein power to a lower baso power, increase in letdown flow rFate, aRd venting of the makeup tank gases to the waste gas deiay tanks.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assUmes specific activity at 1.0 tCi/qm 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 60.0 ujCi/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 2200 ICi/qm DOSE EQUIVALENT XE-1 33.

The SGTR analysis assumes all RCS leakage is initially released via turbine bypass to the condenser. Following a reactor trip, the rise in pressure in the SGs causes radioactively contaminated steam to discharge to the atmosphere through the atmospheric dump valves or the main steam safety valves, which also removes the excess energy to rapidly reduce the RCS pressure and close the main steam safety valves. The ruptured SG is then isolated and the unaffected SG removes core decay heat by venting steam until the event ends when the Decay Heat Removal (DHR) 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 SG removes core decay heat by venting steam to the atmosphere until the cooldown allows the DHR system to be placed in service. The event ends when the RCS temperature reaches 212 OF and no further flashing of any RCS leakage into the affected SG can occur. The SLB is assumed to result in an increase in the total SG tube leakage rate to 1 qpm.

ANO-1 B 3.4.12-3 Amendment No. 215 Rev.

RCS Specific Activity B 3.4.12 Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 60.0 uCi/qm for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

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

The analysis show. the radiologiCal consequonces f _an.SRGTR accideRt a..

within a small fraction of the Reference 1 dose gui!de!lnc i!mits.

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

LCO The iodine specific activity in the reactor coolant is limited to 1.0 wCi/qm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 2200 WCi/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will not exceed the applicable 10 CFR 50.67 requirements (Ref. I ).Th-specific iodwie activity is limited to*

,,3.5 -

Ci/gm DOSE EQUIVALENT 1-131, and th'oa pcfic actiVity in the primary coolant iS limited to the number of ICi~igm equal to 72 divided by E=. The limi4t On DOSE EQUIVALENT!1 131 ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose to an individual at the site bounda~' during the SGTR will be a small fraction o~f the allowcd thyroid dose. The limit en total Specificacivt cnsures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dosc to an individual at the site boundar,' durinth SGTR w..ill be a small fraction of the allow~ed w~hole body dose.

The SLB and SGTR accident analyses 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 applicable 10 CFR 50.67 requirements (Ref. 1 ).Thc analysis shows that the 2 houjr site boundar,' dose levels are withi acetal limits. Violation of the LCO; may result in reactor coo9lant radioEactivity levels that could, in the event of an SGTR, lead to site boun~dar' doses that exceed the 10.CFR 100 dose guideline limits.

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 applicable 10 CFR 50.67 requirements (Ref. 1).In MODES 1 and 2, and in MODE 3 with RCS average tem~peratue Ž! 5002F, operation within the LCOQ limits for DOSE EQUIVALENT 1 131 and total speific activ-ity aFe necessary to limit the potential consequences of a;n SG-T; TR 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 reguired.F-orcertien in MODE 3 with RC r ;vrg t em era 500'F, a*nd in MODES 4 ad 5, the Frelease of rAdnfio;Fcty i, the event f

-n SGTR is unlikely

  • in*e the saturatRon pressure of the ANO-1 B 3.4.12-4 Amendment Ne. 215 Rev.

RCS Specific Activity B 3.4.12 reactor coeoant is below the lift pressure sottings of the atmospheric dump valves and main steam safety valvcs.

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that the specific activity is < 60.0 tJCi/qm. The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide a trend.With the specific aGtiVity of the reactr*

voeaRi greater than Me L6-0.Mi.S, Me. sp.cmTc acviY must be s.o.....

,.0, W

im within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i, adequate to detrmine and imýplement appropriate actions to return specific activity to within limnits.

The DOSE EQUIVALENT 1-131 must be restored-to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 sigqnificant 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it isexpected 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 MODES(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.

C9.1 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 wCi/qm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 challenqing plant systems.ff4he Required Action and associated Completion Time are not met, the reactorF must be ANO-1 B 3.4.12-5 Amendment No. 215 Rev.

RCS Specific Activity B 3.4.12 broughteto MODE 3 with RCS average tempratwr*

500, F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Placing the uit in MODE 3 and RCS average temperature - 500Flwr the saturation pressure of the reactor coolant below the setpOintS Of the mqain steamR safety valves, and prevents venting the SG to the en,,ironment in an SGTR event. The omp'letion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

i required to ro'ch MODE 3 froGm full power conditions in an order!y manner and without challenging unit systemns.

SURVEILLANCE REQUIREMENTS SR 3.4.12.1 SR 3.4.12.1 requires performing a,gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the deqassed 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.SR 3.4.12.1 requires perfori n ma isotopic a*alysis a a mneasue of the gross specifi c activity of the r-eatorycoolant at least once per 7 days.

The geross specific activity analysis cAonsists-o-f the quantitative measur~ement of the total activity of the prima~' coolant in units of mniGrcUriesG "per gram (ICi~igm). Thetoa primar' coolant activity is the sum of the degassed beta gamma activity and the total of all identified gaseous activitis 15 minuLtes after the prim~ar' system is sampled and an" idniidbeta emitters (ietritium, SR89, SR90, etc.). This SuP4eilance pro~vides an indication of an nraein gross specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operatinq conditions. The 7 day Frequency considers the low probability of a gross fuel failure during this time.Tr+eRndi the resuts of this Su.veillance allows proper remedial a*ctio to be taken befor, e reaching the LCO limnit under noFrmal operating conditions. The Surveillance irs applicable in MODES I and 2, and inMODE 3 with RCS average temperature at least 5002F. The 7-cday Fr-eqUeRny isbs the low probability of a gross fuel failue dWring that time pe-rid.

Due to the inherent difficulty in detectinq Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. 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. and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.1242 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 14 day Frequency is adequate to trend changes in the iodine ANO-1 B 3.4.12-6 Am"endment hie. 215 Rev.

RCS Specific Activity B 3.4.12 activity level, considering noble gas activity is monitored every 7 days.This Sur-eillance is pe~forMed in MODE 1 only to cnSUre the iodino rcmnainS within limit during normal operation aRd followffing fastpoWeFr changes when fuel failure i. moe. apt to occur. The 14 day Frequency is adequate to trend changes in t'he8odn act*iviy leVel considering gross secific activity iS monitored every 7 days.

The Note modifies this SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performingq the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

SR 3.4,12.3 rcursrdohemical analysis for E determination every 184 days. The F.

determinatide r

to the LO and is required t verify plant Ope.atiOn within the total specific at;ivity LCO limit. The Frequency of 1814 days recgnizes E does not GhaRge Fap~diy-The radiochomical analysis, conSists of the quantitative measurement of the activity for each radienuclide which is-identPifiAedl in the primar,' coolanti 5 minutes after the primar Gystem is -,ampled.

The acGtiv*imi, for the individual isotepes ae us;,d i the determiRation of E. The gy per disintegration for these radipisotopos detem*i*ed to be present shall be as given in "Tab*p of Isotopes" (1967) (ef 4) and beta energy per disintegration shall be as given in USNRDL= T-R 802 (PartII (Ref. 5) or o)ther references using the equivalent v~alules fo-r the radiG06Gtopes.! Iodine iooi activities are weighted to give DOSE EQUIVALENT 1_131 atfifvit.

This SR is modified by a NOTE that requires the determination be performed w^A+ithin 31 days after iiu of 2 EF=PD and 20 days of MODE 1 operation have elapsed since the re-acGto was, last su-bcritical for at leas,4-8 hours. This ensures the radieactive materials are at equi~libriuim so the analssfrEi rpresentative and not skewed by a crud burs~t or other similar-abno~rm.all event-.

REFERENCES

1.

10 CFR 50.67404.44.

2.

[AST SER reference when received]

2. ANO 1 Operating =icense Amendment 2, (1CGNA057-502) dated May 9, 1975.
3.

10 CFR 50.36.

4.

"Table of Isotopes" (1967).

5.

USNRDL TR 802 (Part I).

ANO-1 B 3.4.12-7 Amendment Ne. 215 Rev.

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