ML061380106

Response to NRC Request for Additional Information on ANO-2 Steam Generator Tube Inservice Inspection Program

ML061380106
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/11/2006
From:	Marlow T A Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN050601
Download: ML061380106 (33)

Text

II I -1. -1 Ah-- Entergy Entergy Operations, Inc.1448 S.R 333 Russellville, AR 72802 Tel 479-8584601 Thomas A. Marlow Director, Nuclear Safety Assurance 2CAN050601 May 11, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

REFERENCE:

License Amendment Request Response to NRC Request for Additional Information on ANO-2 Steam Generator Tube Inservice Inspection Program Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6 1 Entergy letter dated September 19, 2005, Proposed Technical Specification Change to ANO-2 Steam Generator Tube Inservice Inspection Program Using Consolidated Line Item Improvement Process (2CAN090501)

Dear Sir or Madam:

In Reference 1, Entergy Operations, Inc. (Entergy) requested an Operating License amendment for Arkansas Nuclear One, Unit 2 (ANO-2) to replace the existing steam generator tube surveillance program with that being proposed by the Technical Specification Task Force in TSTF 449, Revision 4. TSTF-449, Revision 4 is formatted to the Improved Technical Specification (ITS) plants while the ANO-2 technical specifications (TSs) are based on the CE standard TSs. Therefore, the information contained in TSTF-449, Revision 4 was modified to correspond with the ANO-2 TS format.On March 30, 2006, Entergy received Requests for Additional Information (RAI) from the NRC staff on the proposed amendment request. Attachment 1 provides the response to the requested information.

Attachments 2 and 3 provide markups of the TSs and associated Bases, respectively.

Since several TS and Bases pages are impacted, Entergy is resubmitting markups of all pages associated with the proposed amendment.

The proposed changes to the ANO-2 TSs do not impact the original No Significant Hazards Considerations contained in Reference

1. There are no new commitments associated with this letter.AOL!7 2CAN050601 Page 2 of 3 If you have any questions or require additional information, please contact Steve Bennett at 479-858-4626.

I declare under penalty of perjury that the foregoing is true and correct Executed on May 11, 2006.Sincerely, t A- & u'TAM/sab Attachments:

1. Response to NRC Request for Additional Information for ANO-2 OL Amendment on Steam Generator Tube Inservice Inspection Program 2. Proposed Technical Specification Changes (mark-up)3. Proposed Technical Specification Bases Changes (mark-up)cc: Dr. Bruce S. Mallett 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.O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Drew Holland MS O-7D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205 Attachment 1 2CAN050601 Response to NRC Request for Additional Information for ANO-2 OL Amendment on Steam Generator Tube Inservice Inspection Program Attachment 1 to 2CAN050601 Page 1 of 6 Response to NRC Request for Additional Information for ANO-2 OL Amendment on Steam Generator Tube Inservice Inspection Program NRC RAI 1. Although the staff did not review your cover letter in detail, it did notice that you included a commitment in Attachment 4 that indicates that all loads that can significantly affect burst or collapse will be determined and assessed.

In this commitment, there is a statement that indicates: "These loads, as well as the other analyses to support a 40%plugging limit, will be analyzed for the ANO-2 SG licensing basis. These analyses will be performed and documented under the requirements of 10 CFR 50.59." The NRC staff is aware of the industry's efforts to assess the effects of non-pressure loads on tube integrity (structural and leakage integrity).

These efforts include an assessment of whether changes are needed to the industry guidelines to ensure these loads are appropriately accounted for in tube integrity evaluations (i.e., in the methods used to determine whether the performance criteria have been exceeded).

However, your statements seem to imply that the on-going industry efforts may affect the 40% tube plugging limit. The reason for this is not clear since the 40% plugging limit was developed with consideration of non-pressure loads (consistent with the guidance in Regulatory Guide 1.121). Please clarify the meaning of your commitment which should include a determination of whether it is needed.ANO Response The intent of this commitment is to state that the structural integrity and plugging limit calculation would be completed prior to implementation of the TS amendment.

At the time of the submittal, Entergy had not performed the new structural integrity analysis to comply with NEI 97-06. The new safety factors of 1.2 for primary loads and the 1.0 for secondary axial loads have to be added to the current analysis for tube plugging.

There is no change to the 40% tube plugging limit. It is believed that the analysis results can be incorporated into the ANO-2 licensing basis under IOCFR50.59 and should not require NRC review and approval.NRC RAI 2. In the ACTION section of TS 3.4.5, "Steam Generator (SG) Tube Integrity," paragraph a.1 of your proposal indicates: "Within 7 days verify tube integrity of the affected tube(s) is maintained until the next inspection, and." For the same section, TSTF-449 indicates:

VWithin 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection." The underlined words (which are part of the TSTF) could eliminate the need to shutdown the facility in the event that tube integrity is only maintained until a refueling outage and not until the next SG tube inspection.

Please discuss your plans to revise your proposed TS to make them consistent with the TSTF.

Attachment 1 to 2CAN050601 Page 2 of 6 ANO Response Entergy agrees with the identified change. Action a.1 for TS 3.4.5 will be revised to: Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection.

The revised TS page is contained in Attachment 2.NRC RAI 3. In the ACTION section of TS 3.4.5, "Steam Generator (SG) Tube Integrity," paragraph b appears to have a typographical error. It appears that "with" should be"within".

Please discuss your plans to modify your proposal to address this comment.ANO Response Entergy agrees, Action 3.4.5.b should read:... and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> [emphasis added].The revised TS page is contained in Attachment 2.NRC RAI 4. In your proposed surveillance requirement 4.4.6.2.2 for Reactor Coolant System Leakage (page 3/4-14 has "Reactor Coolant System Operational Leakage" and page 3/4-14a has "Reactor Coolant System Leakage"), the primary-to-secondary limit is referenced as 150 gallons per day Der SG. This wording is inconsistent with TSTF-449 and the wording in your proposed 3.4.6.2.c in which the limit is referenced as 150 gallons per day through any one steam generator.

Please discuss your plans for modifying your proposed TS to make the wording consistent with the TSTF.ANO Response Entergy will correct the SR 4.4.6.2.2 to read: Primary to secondary leakage shall be verified to be 150 gallons per day through any one SG at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> [emphasis added].The revised TS page is contained in Attachment 2.NRC RAI 5. Proposed TS section 6.5.9.a appears to have some typographical errors. In the last phrase of the third sentence, it appears that it should read "prior to the plugging of tubes". In addition, it appears that the last sentence should read, in part "....are inspected or plugged....." This latter typographical error is also a typographical error in the TSTF.Please discuss your plans to modify your proposal to address these comments.ANO Response Entergy agrees with the identified corrections in TS 6.5.9.a. The revised TS page is contained in Attachment

2.

Attachment 1 to 2CAN050601 Page 3 of 6 NRC RAI 6. Proposed TS 6.5.9.d is not fully consistent with TSTF-449.

In particular, the entire length of tube is not specified (i.e., "from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet") and the objective of the inspections is not fully specified

("and that may satisfy the applicable tube repair criteria").

In addition, your proposal does not include the following from TSTF-449 (section 5.5.9.d.1): "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement." The staff notes that this will avoid future modification of your TS in the event that another SG replacement is needed.Please discuss your plans to modify your proposal to make it consistent with the TSTF.The staff notes that if section 5.5.9.d.1 from TSTF 449 is included in your proposal, the body of your proposed 6.5.9.d will also need to be revised to reference sections, "d.1, d.2, and d.3 below." ANO Response The second sentence of TS 6.5.9.d will be revised to read: The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.The following action will be added to TS 6.5.9.d: as Action 1. The subsequent actions will be renumbered as Actions 2 and 3 and properly referenced in the previous paragraph of TS 6.5.9.d.1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

The revised TS page is contained in Attachment 2.NRC RAI 7. Proposed TS 6.5.9.b.3 refers to LCO 3.4.6.2 as "RCS Operational Leakage." The title of LCO 3.4.6.2 in your technical specifications is either "Reactor Coolant System Operational Leakage" or "Reactor Coolant System Leakage" (depending on the page).Please discuss your plans to make these references consistent.

ANO Response"Reactor Coolant System" is used in ANO-2 TS LCO 3.4.6.2 when referring to operational leakage. Therefore, Entergy will correct the reference in TS 6.5.9.b.3 and TS 4.4.6.2.1 to read of "Reactor Coolant System operational leakage." Similar corrections are being made to TS Bases 3/4.4.5, where appropriate.

The revised TS pages are contained in Attachment 2 and the revised TS Bases pages are contained in Attachment

3.

Attachment 1 to 2CAN050601 Page 4 of 6 NRC RAI 8. In your submittal, you did not provide any changes to the BASES section dealing with Reactor Coolant System Operational Leakage requirements (except for inclusion of insert B 3.4.13B from TSTF-449).

Please discuss whether the changes indicated in the TSTF were (or will be) made to the appropriate BASES sections.ANO Response The Bases for the ANO-2 TSs are based on the CE standard TSs and do not contain the same level of detail as improved standard TSs. Therefore, the detail provided in TSTF-449 may not be necessary for inclusion in the ANO-2 TSs unless it provides additional insights to the TSs. Notwithstanding, Entergy proposes to add the following additional information into the ANO-2 TS Bases 3/4.4.6.2 consistent with TSTF-449: The 150 gallons per day limit is measured at room temperature as described in EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

The operational leakage rate limit applies to leakage through any one SG. If it is not practical to assign the leakage to an individual SG, all the primary to secondary leakage should be conservatively assumed to be from one SG.For primary to secondary leakage determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. The surveillance frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents.

The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

The revised TS Bases page is contained in Attachment 3.NRC RAI 9. In the Background section of BASES Section 3/4.4.5, "STEAM GENERATOR (SG) TUBE INTEGRITY", the first paragraph leaves out the last four sentences from the corresponding paragraph in TSTF-449 which indicates:

The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4,"RCS Loops -MODES I and 2," LCO 3.4.5, 'RCS Loops -MODE 3," LCO 3.4.6, "RCS Loops -MODE 4," and LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled." Although the staff recognizes that some of the section numbers may be different at ANO-2, it is not clear why these sentences were omitted. Please discuss your plans for incorporating these sentences (with the appropriate section numbers) into your BASES.

Attachment 1 to 2CAN050601 Page 5 of 6 ANO Response Entergy agrees to add the TSTF-449 discussion to the ANO-2 Bases (modified for ANO-2 TSs).The SG tubes isolate the radioactive fission products in the prnmary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.1.1, STARTUP and POWER OPERATION (MODES 1 and 2), LCO 3.4.1.2, HOT STANDBY (MODE 3), and LCO 3.4.1.3, SHUTDOWN (MODES 4 and 5)The revised TS Bases page is contained in Attachment 3.NRC RAI 10. In the Limiting Condition for Operation section of your BASES Section 3/4.4.5,"STEAM GENERATOR (SG) TUBE INTEGRITY", the reference to Regulatory Guide 1.121 is omitted from the bullet dealing with the structural integrity performance criterion (i.e., where Subsection NB of the ASME Code is referenced).

Since Regulatory Guide 1.121 was used in the development of the structural integrity performance criterion, it is not clear why it is not referenced.

Please discuss your plans to modify your proposal to address this comment.In the same section, the operational leakage performance criterion is referenced as 150 gallons per day per SG. Since this sentence already indicates that the leakage is through any one steam generator, it is not clear why the 'per SG" is included.

See related question pertaining to this issue. Please discuss your plans to modify your proposal to make it consistent with TSTF-449 (which does not include the 'per SG").ANO Response The new structural integrity analysis that is being performed for ANO-2 includes the analysis performed per draft Regulatory Guide 1.121 and therefore was not initially included.

However, since the structural integrity analysis incorporates approaches and methodologies from Regulatory Guide 1.121, the reference to draft Regulatory Guide 1.121 will be added to the Bases.Entergy agrees that the "per SG" in the LCO Bases to 3/4.4.5 is redundant and should be removed. Additionally, the previously proposed change to TS Bases 3/4.4.6.2 which referred to "150 gallons per day per SG" will be revised to "150 gallons per day through any one SG". Even though this is different than TSTF-449, the change will be consistent with other proposed changes.The revised Bases pages are contained in Attachment

3.

Attachment 1 to 2CAN050601 Page 6 of 6 NRC RAI 11. In the Surveillance Requirement section of BASES Section 3/4.4.5, "STEAM GENERATOR (SG) TUBE INTEGRITY', the last paragraph from TSTF-449 on the frequency of the surveillance was not included.

The reason for the exclusion is not clear.Please discuss whether the paragraph from TSTF-449 will be added to your BASES section.ANO Response The ANO-2 TS Bases 3/4.4.5 will be revised to add the following:

The Frequency in SR 4.4.5.2 of "prior to entering HOT SHUTDOWN following a SG inspection" ensures that the SR has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

The revised Bases page is contained in Attachment 3.Additional Proposed Changes to ANO-2 TSs for Consistency with TSTF-449 1. In subsequent review of the ANO-2 TSs, Entergy identified that a note had not been included with the initially proposed TS change. TSTF-449 proposed a Note identified as Insert 3.4.13A which is applicable to TSTF-449 SRs 3.4.13.1 and 3.4.13.2.

This note reads Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Entergy is adding this note to the proposed changes for ANO-2 SRs 4.4.6.2.1.a and 4.4.6.2.2 consistent with TSTF-449.

The corrected page is contained in Attachment

2.2. Several

locations within the proposed TSs when referring to Hot Standby, Hot Shutdown and Cold Shutdown are "defined terms" but were not capitalized as such. These terms have been corrected to HOT STANDBY, HOT SHUTDOWN and COLD SHUTDOWN, as appropriate.

3. A clarification is being made to the Bases of TS 3/4.4.5, Safety Analysis, to state that the described SGTR accident analysis is under the assumption that offsite power is available.

A reference to ANO-2 SAR Section 15.1.18 is also proposed.

Attachment 2 2CAN050601 Proposed Technical Specification Changes (mark-up)

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be: a. Analog channels -The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels -The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c. Digital computer channels -The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation 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 ALTERATION shall not preclude completion of movement of a component to a safe conservative position.SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: a. Leakage (except controlled leakage) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or b. 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 c. Reactor coolant system leakage through a steam generator to the secondary system (primarv to secondary leakage).ARKANSAS -UNIT 2 1-3 Amendment No. 415,220, 2. I DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam ge;%rate tube rimarv to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.AZIMUTHAL POWER TILT -Tg 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT 1-131 1.18 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (p.Ci/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table IlIl of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." E -AVERAGE DISINTEGRATION ENERGY 1.19 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of: a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.ARKANSAS -UNIT 2 1-4 Amendment No. 157, 255, REACTOR COOLANT SYSTEM STEAM GENERATORS (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Each eteam generate hafll be OPERA.BLE.

a. SG tube integrity shall be maintained.

b. All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With one or mere Eteam generatore inoperable, restore the inoperable gencrator(S) to OPERABLE status prior to increasing Thvg above 200 0 F.The Actions may be entered separately for each SG tube.a. With one or more SG tubes satisfying the tube repair criteria and are not plugged in accordance with the Steam Generator Program.1. Within 7 days verify tube intearity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and 2. Plug the affected tube(s) in accordance with the Steam Generator Program porior to entering HOT SHUTDOWN following the next refuel na outage or SG tube inspection

b. If the reguired Action and Allowed Outage Time of Action a. above cannot be met or the SG tube integrity cannot be maintained, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE REQUIREMENTS 4.4.5.1 Each steam generator chall bo demonstrated OPERABLE in accordance with the Steam Generator Tube Surveillance Program. Verify SG tube integrity in accordance with the Steam Generator Program.4.4.5.2 Verify that each Inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

ARKANSAS -UNIT 2 3/4 4-6 Next Page is 3/4 4-13 Amendment No. 468,48,249,217,223, 233,255, REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to: a. No PRESSURE BOUNDARY LEAKAGE, b. 1 GPM UNIDENTIFIED LEAKAGE, c. 300 gallons p d t primary to secondary leakage thrugh both steam generators and 150 gallons per day Primary to secondary leakage through any one steam generatorsGI, d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e. Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6.1.APPLICABILITY:

MODES 1,2, 3 and 4.ACTION: a. With any PRESSURE BOUNDARY LEAKAGE or any primarv to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and orimarv to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two valves* in each high pressure line having a non-functional valve and be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.* These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.ARKANSAS -UNIT 2 3/4 4-14 Amendment No. 484, Ordetzr datedr' 4120121 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages.

except for Primary to secondarlleakaae. shall be demonstrated to be within each of the above limits by: a. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation except when operating in the shutdown cooling mode*.b. Monitoring the reactor head flange leakoff temperature at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.4.4.6.2.2 Primary to secondary leakage shall be verified to be 150 gallons per day through any one SG at least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s*.4.4.6.2.23 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4.6-1 shall be demonstrated OPERABLE by individually verifying leakage to be within its limit: a. Prior to entering MODE 2 after each refueling outage, b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months, and c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.* Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

ARKANSAS -UNIT 2 3/4 4-14a Order date 1/20/81 Amendment No. 234, ADMINISTRATIVE CONTROLS 6.5.9 Steam Generator (SG) Tube Surcillanc Program A Steam Generator Proaram shall be established and implemented to ensure that SG tube integrity is maintained.

In addition.

the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the 'as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice Inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, or pluaged to confirm that the performance criteria are being met.b. Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.1 .Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup. operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident Primary to secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1,2 on the combined Primary loads and 1.0 on axial secondary loads.2. Accident induced leakage oerformance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture.shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed I opm through any one SG.3L The operational leakage nerformance criterion Is specified in LCO 3.4.6.2. uReactor Coolant System operational leakage." c. Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be pluaged.6.5.9.1 Steam Generator Sapnrle Seletinof-n and Inp ARKANSAS -UNIT 2 6-8 Amendment No. 2-55, Each steam gencrator shall be determined OPERABLE during shutdown by selecting and iRsperting at Ieast the rnirmum RumFor oF etfseam g t pecfied On Table 6.5.0 1t.6.5 .02 $team GeF^erat Tu Sample Seyemtian ar-- -*-* e .k The tear genoratr tube mninimum sample size, inspection result classification, and tho corGcsponding action required shall be as specified in Table 6.5.0 2. The inscrpvie inspection of steam goeerator tubes shall be performed at the frequencies specified In Specification 6.5.9.3 aRd the inSpected tubes shall be w.erified acceptable per the -aecptanec criteria of SpecificatioR 6.5.OA4. The tubes selected for each insr .ice inspection shall include at least 3% cf the total Rumber of tubes in; all steam generatrs; the tubes selerted for these irspections shall be soleeted on a random basis exeept: a. Where experiece in similar plants with similar water chemistry indicates critical areas to be InSpected, thcn at Icast 50% of the tubes inspccted shall be from these critical areas.b. The first sample of tubes selected for each inservice inspection (subsequent to the pre ervice insection)f eh steam generoateor shagl inclue 1. All non plugged tubes that previously had detectable wall penetrations ( 20%).2. Tubes in those areas where experience has indicated potential problems.3. A tube inspection (pursuant to Specification 6.5.9.4.a.9) shall be performed on each elected tubte. If any selected tube does Rot permit tha ge of the eddy curreRt probe foF a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c. The tubes selected as the second and third samples (if required by Table 6.5.9 2) during cash inseor'ie inspecton may be subjected to a patial inspeton;R provided: 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.2. Thc inspections include those porions of the tubes where imperfections were previously found.ARKANSAS -UNIT 2 6-8 Amendment No. 255, ADMINISTRATIVE CONTROLS ADMINISTRATIVE CONTROLS d. Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be Performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet.and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not Part of the tube. In addition to meeting the reguirements of d.1 d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be emploved and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144. 108. 72. and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less)without being inspected.1

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s).

then the indication need not be treated as a crack.e, Provisions for monitoring operational Primarv to secondary leakage.The r^eult of each sample inspertion shall beo lassified Into ono of the following three Catege Results Loss than 5% of the total tubes- inspected are degraded tubes and none of the inrspeted tubes arc defective.

One or more tubes, but not more than 1% of the total tubes inspected are defcstive, or between 5% an 10% of the total tubes inspected arc degraded tubes.6-4 3More than 10% of the total tubes inspected are degraded tubes or more than 1°/ of the inspcsted ARKANSAS -UNIT 2 6-9 Amendment No. 2 Next Paae is 6-14 tubes are defective.

Note: In all inspections, previously degi eignifitant

(- I0%) n vrthor wall pe above percentage calculations.

em led tubes must exhibit otrations to be included 8n the ARKANSAS -UNIT 2 6-9 Next Pace is 6-14 Amendment No. 26 I ADMINISTR, TIVE CONTROLS 6.5.9.3 Insgrctioen Frepuoncios The above required inser'ice inspections of steam gencrmtor tubes shall bc performed at the folloWin; frequencies:

a. The first n cinspetn shall be performed after 6 Merffti'e Full P. weAMonths but.within °1 calendar monthM of initial eriticality.

Subsequent inscrviccinspections shall be porformed at intervals of not less than 12 nor more than=24 calendar months aftcr the pevious inspection.

If two conseeutive

nspecti;ns following-i under AVT conditions, not including the pre service inspection,rcsult in all inspection results falling into the C I Gatcgory or if tuivejnspoetins demonstrate that prev;ously obscr.'ad degradation has net continuedzand no additional degradation has occurred, the inspection intcrval may bce ended to a maximum of once per 10 months.A one time inspection interal of a maximum of once per 40 months is allowed for the inspertion performed immediately foloGw;ig the 2R1 5 outage. This ir an exccption to 6.5.9.3.a in that the interval extension is based on all of the results of on inspection.

falling into the C I4 Gatcgry.b. If the results of the iRsevr;e inrspection of a steam generator onRducted

n.Raceedance with Table 6.5.u 2 at 40 month inter.als fall into Category C 3, thcinspection frequency chall be increased to at least once per 20 months. ThcRncrease in inspection frequency shall apply until the subsequent inspoctinssatisfy the criteria of SpeGificatioR C5I;the interval may then be exended toa maximum of once per 10 months.c. Additional, unscheduled inecrce inspectiones shall be performed on each steam generator in accordance with the first sample inspection specified in Table 6.5.0 2 during the shutdown subsequent to any of the fllowin;g condition;s

1. Pldmar, to seondWa tube leaks (not incmdLng leaks efkginaft f_ ngue te tube sheet wklds) in excess of the limit& of Specification

34. 6.2.2. A seismi m ocureRne greater thaR the Operating Basis Eathquake.

_. A lossofncol a Vnt ccd7ent requiring a i o e gimeeFed 4. A m~ain steam !ine or feedwater lie break-.ARKANSAS UNIT 2 6 10 Amendment No. 255 I ADMINISTRATIVE CONTROLS.. .Ac.eiqtance C r.t.ria a. As used in this Specification

1. Tubina orFTube mcnans that portionof the tubcewhich forms the.pimr cystem to ewcndry system prcSSrc boundary, 2. Imnerfection means an exception to the dimensions, finish or contour of a tube froe that required by fabrioation dJrawings or speocifications.

Eddy current testing indications below 20% of the nominal tube wall thickness, if detectablc, may be enRsidered as imperfons.

.Degradation rne; 1-ond le G~or~ldng wstaget^ waF eF genersarl corrosion occurring on either inside or outside of a tube.4. Deraded Tube means a tube ontaining impeerfctilens 20% of Rnminal wall thickness caused by degradation.

I; 0/ Dearadatiorn rrean theZ pementragte efthe tube wall1 thiorknessae ofed or removed by degradation.

6. Defect means an impefection of such severity that it exceeds the plugging limit. A tube containing a defect ic defcrtive.

7. Plugging 6imfit means the imp.fvo depth at or beyond which the tube shall be removed from Sor^ice by prlugging because it may become unser:icoable prior to the next inspection.

The plugging limit is equal to 40% of the nominal tube wall thickness.

Q. Urieabedescribes thecndition of .a tube ffif it leaks Or con-;tain a defecat large enough to affect fts strFutural irtegrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a steam line or feedwater lin bra sspecified in 6.5.19.3.c, above-.Q. Tube Inesection means an inspection of the steam generator tube from tube end.(eold leg side) to tube end (hot leg side)10. Pro service Inencstion means an inspection of the full length of each tube in each steam generator performed by eddy current tcchniques prior to ccrvice to cstablish a baseline condition of the tubing. This inspection shall be performed after the hydroestati test and prior to PO'NER OPERAJTION using the equipment and techniques expected to be used during subs queni vire inspetions.

b.v The steamFenertr shall be determined OPERA.BLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes centaining through wall cracks) required by Table 6.56. 2.A OA IA^ I IL, 1, #^* A A idI o1 i i A;~ecnumcnefi I'Eo.I ADMINISTRATIVE CONTROLS TABL=E-6.59 i MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Pre scevice Inspection No. of Steam Generators per Unit Twe Flrst Inscr.vicc Inspection One Second & Subsequent Inccrsiee Inspections Qpe'Thbl Notation:.4 'rk- :- ----- L.-- --.. 6. --- --- -- ---L.-A. I-* * .. ** **

• n, ,,-- * .... s. *1 i~nrnmni~i 2 Ng 04 a~ thP ti th~r, (AAFhACP~

NI ; thcp ni imhpr nf gtnizm rnptrrmtnr in~ thim plant) if the results of the fsrct or previous inspections indicatc that all steam gencrators are pe.rforming In a like manner. Note that under some circumstances, the operating Genditions in onc or more steam generators may be found to be more severe than those in ether steam generators.

Undcr such circumstances the sample sequence shall be modified to inspectthe most severe conditions.

A EIPA If% A^ a I kI iIr.A ~ -_.L.- ^rrLJrI I fl I -"r' I ADMINISTRATIVE CONTROLS TABLE 6.5.09 2 STEAM GENERATOR TUBE INSPECTION 41TM PEINSPEN 2ND SAMPLE INSPECTiON l 3RDSAMPLE INSPECTION

_apl Acsu4 Action Reqie o eu GG Requie eu ~ on eud A G4 Noe NA NWA WA NA p _ _ __ __ _ANIA NA C4 Plu.g dofectivo tubes G4 PIGA8 N/A NWA and inspect additio"al

_2S tubes in thie S.G. Plug defoctiv tubes G4 Nope G42 and ieetG2 Plug defefoci~:

u additio2 n = S tub4 R Perform Fction for C 3 G3 f _ G4 roWult of firNt eamp Porform ecton for _G4 C3 rosultof first N/A N/A_ __6am0 _G4 Inepoct all tubee in Qthev thsS.G.plug

&is Nene W/A N/A doet tubes and G4 Inspect 2S tube& in&etheA G Other Pefrm action for N/A NA&Ga-or C 2 Fesiflt of se~ood Speooal DC-2 sample NRC per Speficaton Other Inopct all tuboc in S.Gi the other S.G. and G4 plug defectivc tubos.N/A N/A Sposhal Roport to NRC per Spec. 6.6.7 S -3 (2/n) % Where n is the number of stcam generators inspected during an inspection.

ARKAINSAS UNIT 2 6 13 Amendmpnt No. 25S I ADMINISTRATIVE CONTROLS 6.6.6 not used 6.6.7 Steam Generator Tube Surveillance-Inspection Reports A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed In accordance with the Specification 6.5.9, Steam Generator (SG) Program. The report shall include: a, The scone of inspections performed on each SG.b. Active degradation mechanisms found.C. Nondestructive examination techniques utilized for each degradation mechanism.

_. Location.

orientation (if linear), and measured sizes (if available) of service induced indications.

e. Number of tubes plugged during the inspection outage for each active degradation mechanism.

f. Total number and percentage of tubes plugged to date.g. The results of condition monitoring, including the results of tube pulls and in-situ testing.h. The effective plugging percentage for all plugging in each SG a. Following each inscr'Ace inspection of steam generator tubes the number of tubes plugged in each steam generator ehall be reparted to the Commeissin within 45 days.b. The complete results of the steam generator tube inser'.'iee inspection shall be repoFAtd within 12 months following the completion of the inco )A in en.This report shall include: 1. N~umber and exent of tubes inspected.

2. Location and percent of wall thickness penetration for each indication of an ipeteren?3. Identification of tubes plugged.c. Results of steam generator tube inspections which fall into Category C 3 shall be repoeted to the ted by Table 6.5.0 2. Notifir-clafilon ef the Commission will be made prior to resumption of plant operation (i.e., prior to entering Mode 4). The written report shall provide a description of investigations conducted to determin; cause o the tube degradation and corrective measures taken to prevent recurrence.

6.6.8. Specific

Activity ARKANSAS -UNIT 2 6-22 Amendment No. 255, 257, The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded;(2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded the results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.ARKANSAS -UNIT 2 6-22 Amendment No. 255, 257, Attachment 3 2CAN050601 Proposed Technical Specification Bases Changes (mark-up)

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves against water relief. The steam bubble functions to relieve RCS pressure during all design transients.

The requirement that 150 KW of pressurizer proportional heaters per bank and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss-of-offsite power condition to maintain natural circulation at HOT STANDBY. Action (b) is applicable to conditions when a single proportional heater bank is inoperable for any reason, whether loss of an emergency power supply or the loss of capability (less than 150 KW output per bank as required by Surveillance 4.4.4.2(b)).

Action (b) requires that the AOT associated with a single inoperable EDG (TS 3.8.1.1 Action b.3) be entered when a heater bank has an inoperable emergency power supply. This action allows 14 days to restore an inoperable EDG to OPERABLE status provided the AACDG is available.

If the AACDG is not available, the EDG must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In the event the circuit between the AACDG and the proportional heater bank is unavailable (480 V Load Center breaker supplying the heater bank is open or other similar break in the circuit), the AACDG cannot be considered available to the heater bank. Likewise, if a proportional heater bank output capability is known to be < 150 KW, the availability of the AACDG provides no additional capability.

Both of these cases require the proportional heater bank to be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in accordance with Note 1 of EDG TS 3.8.1.1 Action b.3 (Reference Licensing Memo LIC 04-045 and -046).3/4.4.5 STEAM GENERATORS (SG) TUBE INTEGRITY The Steam uenert Tbe Su rlllaee PDrogram ensures that the set al integrty f this portion of the RCS will be maintained.

The program is based on a modification of Regulatory Guide 1.83, Revision 1. Inser.vice inspection of steam generator tubing is essential in order to ma_;tain GU ...lllance of the conditieos of the tubes in_ the evenrt that there is evidence mechanical damage or prgreessire degradaton due to design, or inseNrico onditions that lead to corroslon.

insor ieo inspection of steam generator tubing also per-evide a meansw ef Gha~rtezngo;^^

the natwe ar d Gaurl s eo ef any tub slot ^adaltiern so th corrective measures can be taken.Backoround Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory.

The SG tubes isolate the radioactive fission products in the Rrimary coolant from the secondary system. In addition, as Dart of the RCPB. the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the Primarv system. This Specification addresses only the RCPB intearitv function of the SG. The SG heat removal function Is addressed by LCO 3.4.1.1. STARTUP and POWER OPERATION (MODES 1 and 2), LCO 3.4.1.2. HOT STANDBY (MODE 3). and LCO 3.4.1.3. SHUTDOWN (MODES 4 and 5)ARKANSAS -UNIT 2 B 3/4 4-2 Amendment No. 20,158,184,223 Rev. 4G,44, SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis. including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms.

Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting. interaranular attack, and stress corrosion cracking.

along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube Integrity if they are not managed effectively.

The SG Performance criteria are used to manage SG tube degradation.

Specification 6.5.9. "Steam Generator Program." requires that a program be established and implemented to ensure that SG tube integrity is maintained.

Pursuant to Specification 6.5.9.tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria:

structural integrity, accident induced leakage, and operational leakage.The SG performance criteria are described in Specification

6.5.9. Meeting

the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by NEI 97-06, Steam Generator Program Guidelines (Ref. 1j)Safety Analysis The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding a SGTR is the basis for this Specification.

The analysis of a SGTR event assumes a bounding Primary to secondary leakage rate eaual to the operational leakage rate limits in LCO 3.4.6.2. "Reactor Coolant System Operational Leakage." olus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR when offsite power Is available assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

See ANO-2 SAR Section 15.1.18.The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e.. they are assumed not to rupture.)

In these analyses, the steam discharge to the atmosphere is based on the total Primary to secondary leakage from all SGs of 1 gallon per minute or is assumed to increase to I gallon ner minute as a result of accident induced conditions.

For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.8. "Reactor Coolant System Specific Activity." limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g.. a small fraction of these limits).Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation The LCO requires that SG tube integrity be maintained.

The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Prosram. During a SG inspection.

any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall.between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.ARKANSAS -UNIT 2 B 3/4 4-2 Amcndment No. 20,158,181,223 Rev. 40,44, A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification

6.5.9. Steam

Generator Proaram. and describe acceptable SG tube performance.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.There are three SG oerformance criteria:

structural integrity, accident induced leakage. and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.* The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification.

Tube burst is defined as. "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (olastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure.

collapse occurs at the ton of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that significantly affect burst or collapse.

In that context, the term"significantly" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations.

except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between Primary and secondary classifications will be based on detailed analysis and/or testing.Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code. Section IlIl. Service Level A (normal operating conditions) and Service Level B (unset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code. Section 111. Subsection NB and draft Regulatory Guide 1.121.* The accident induced leakage perfornmance criterion ensures that the Primarv to secondary leakaae caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions.

The accident analysis assumes that accident induced leakage does not exceed 1 qpm through any one SG. The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to Primarv to secondary leakage induced during the accident.* The operational leakage Performance criterion provides an observable indication of SG tube conditions during plant operation.

The limit on operational leakaae Is contained in LCO 3.4.6.2. Reactor Coolant System Operational Leakacie.

and limits Primary to secondarv leakage through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

Actions ARKANSAS -UNIT 2 B 3/4 4-2 Amcndmcnt No. 20,158,181,223 Rev. 4X,44, The Actions may be entered separately for each SG tube. This is acceptable because the Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Actions may allow for continued operations, and subsequent affected SG tubes are governed by subsequent application of associated Actions.Action "a" applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.4.5.2. An evaluation of SG tube integritv of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube Integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection.

The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection.

If it is determined that tube integrity is not being maintained.

Action Wb" applies.An allowed outage time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity.

Action a.2 allows Plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.However, the affected tube(s) must be plugged prior to entering HOT SHUTDOWN following the next refueling outage or SG inspection.

This time period is acceptable since operation until the next inspection is supported by the operational assessment.

Action "b" applies if the actions and associated allowed outage time of Action 'a" are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed outage time are reasonable, based on operating experience.

to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.Surveillance Requirements During shutdown periods the SGs are inspected as required by SR 4.4.5.1 and the Steam Generator Program. NEI 97-06. Steam Generator Proqram Guidelines (Ref. 1). and its referenced EPRI Guidelines, establish the content of the Steam Generator Proaram. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed.

The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.The Steam Generator Program determines the scone of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria.

Inspection scope (i.e.. which tubes or areas of tubing within the SG are to be Inspected)

Is a function of existing and potential degradation locations.

The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function ARKANSAS -UNIT 2 B 3/4 4-2 Amcndment No. 20,158,184,223 Rev. 40,44, of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequencv is determined by the operational assessment and other limits In the SG examination guidelines (Ref. 6). The Steam Generator Proram uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection.

In addition.Specification

6.5.9 contains

prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

As required by SR 4.4.5.2. any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

The tube repair criteria delineated in Specification 6.5.9 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subiect tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency in SR 4.4.5.2 of uPrior to entering HOT SHUTDOWN following a SG inspection" ensures that the SR has been completed and all tubes meeting the repair criteria are plugged prior to subiecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES

1. NEI 97-06. "Steam Generator Program Guidelines." 2. 10 CFR 50 Appendix A. GDC 19.3. 10 CFR 100.4. ASME Boiler and Pressure Vessel Code. Section Ill. Subsection NB.5. Draft Regulatory Guide 1.121. "Basis for Plugging Degraded Steam Generator Tubes," August 1976.6. EPRI. "Pressurized Water Reactor Steam Generator Examination Guidelines." ARKANSAS -UNIT 2 B 3/4 4-2 Amendmct No. 20,158,184,223 Rev. 40,X4, REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.The total steam genertor t Icakage limit of 300 gallons per day for all steam genr atoa ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of ethter a steam generator tube rupture or steam line break. The 150 gallon per day leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

The limit of 150 gallons per day through any one SG is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines which states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons Per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in coniunction with the implementation of the Steam Generator Program is an effective measure for minimizing the freauencv of steam generator tube ruptures.The 150 gallons per day limit is measured at room temperature as described in EPRI, Pressurized Water Reactor Primarv-to-Secondart Leak Guidelines.

The operational leakage rate limit applies to leakage through any one SG. If it is not practical to assign the leakage to an individual SG. all the primary to secondarv leakage should be conservatively assumed to be from one SG.For Primary to secondary leakage determination, steady state is defined as stable RCS pressure.

temperature, power level, pressurizer and makeup tank levels, makeup and letdown.and RCP seal injection and return flows. The surveillance frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a Leasonable interval to trend primary to secondary leakage and recognizes the Importance of early leakage detection in the prevention of accidents.

The Primarv to secondarv leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of ARKANSAS -UNIT 2 B 3/4 4-2 Amendment No. 20,158,184,223 Rev. 4X,14, any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a ARKANSAS -UNIT 2 B 3/4 4-2 Amrendmcnt No. 20,158,184,223 Rev. 4X,44,