2CAN120401, License Amendment Request Proposed Technical Specification Changes Revising Containment Building Structural Integrity Requirements

License Amendment Request Proposed Technical Specification Changes Revising Containment Building Structural Integrity Requirements

ML050130173
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/20/2004
From:	Forbes J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN120401
Download: ML050130173 (18)

Text

I Entergy Operations, Inc.

6Entergy 1448SR 333 Russellville, AR 72802 Tel 4794584888 Jeffery S. Forbes Vice President Operations ANO 2CAN120401 December 20, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Proposed Technical Specification Changes Revising Containment Building Structural Integrity Requirements Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

REFERENCES:

1 Letter from NRC to Entergy dated September 9, 1999, Arkansas Nuclear One, Unit No. 1, Issuance Of Amendment Re: Reactor Building Structural Integrity Surveillance Requirements (1CNA099901)

Dear Sir or Madam:

Attached for your review and approval are proposed Technical Specification (TS) changes revising the requirements associated with Arkansas Nuclear One, Unit 2 (ANO-2) provisions for containment building testing and inspection. The proposed changes affect ANO-2 TS Limiting Conditions for Operation (LCO), Surveillance Requirements, and applicable Bases relevant to inservice inspection requirements for the containment structures and tendons. Specifically, ANO-2 has implemented a containment inspection program which is in compliance with the requirements of 10 CFR 50.55a. The program is based on the American Society of Mechanical Engineers (ASME) Section Xl, Subsection IWL, as required by 10 CFR 50.55a(b)(2). Therefore, the proposed changes are being made to update the ANO-2 TSs to current requirements and format.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in Attachment 1. There are no commitments being made as a result of this request.

2CAN120401 Page 2 of 2 The proposed change is neither exigent nor emergency however; your prompt review is requested. We request NRC approval by August 31, 2004. Once approved, the amendment shall be implemented within 90 days. If you have any questions or require additional information, please contact Steve A Bennett at 479-858-4626.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 20, 2004.

Sincerely, JSF/sab Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Proposed Technical Specification Bases Changes (mark-up). For Information Only 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. Box310 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 I 2CAN1 20401 Analysis of Proposed Technical Specification Change to 2CAN120401 Page 1 of 6 Analysis of Proposed Technical Specification Change

1.0 DESCRIPTION

OF PROPOSED CHANGES The proposed change revises the Technical Specifications (TS) for Arkansas Nuclear One, Unit 2 (ANO-2). The proposed changes modify the Containment Structural Integrity specification (TS 3.6.1.5) to delete the existing Surveillance Requirements (SR) and add a new SR to verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. A new Containment Tendon Surveillance Program requirement is being added to TS 6.5.6 and a new reporting requirement is being added to TS 6.6.6. The proposed changes are generally consistent with NUREG 1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.

2.0 PROPOSED CHANGE

The proposed changes to the Arkansas Nuclear One, Unit 2 TS are as follows:

• The Surveillance Requirement for TS 3.6.1.5 has been modified to delete the previous structural integrity Limiting Condition for Operation (LCO) requirement and to add a new LCO to ensure the structural integrity of the containment shall be OPERABLE. The Action statement is modified to be consistent with the LCO. The existing SRs are being replaced with a single SR which states: Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program.

• An administrative change is being made to TS page 3/4 6-9a to renumber it as page 3/4 6-9.

• A new Containment Tendon Surveillance Program is being added as TS 6.5.6 which states:

This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The Containment Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with the ASME Code, Section Xl, Subsection IWL and 10 CFR 50.55a. The provisions of SR 4.0.3 are applicable to the Containment Tendon Surveillance Program inspection frequencies.

This change meets the intent of TS 5.5.6 of NUREG-1432, Revision 3 but was modified as follows:

1. Regulatory Guide (RG) 1.35 is referenced in NUREG-1432, however, it has been superseded by 10 CFR 50.55a and ASME Code Section IWL for containment inspections. This is further discussed in Section 4.0 of this attachment.

2. The sentence regarding the program shall include baseline measurements prior to initial operations was deleted. This statement is no longer applicable to ANO-2 containment inspections.

3. The provisions of SR 4.0.2 being applicable are not being included in TS 5.5.6.

As discussed in the TS Bases of SR 3.0.2 of NUREG-1432, Revision 3, the requirements of regulations take precedence over the TSs. Therefore, SR 4.0.2 of the ANO-2 TSs is not applicable when referencing 10 CFR 50.55a as the inspection requirements.

to 2CAN120401 Page 2 of 6

• An administrative change is being made to TS 6.5.7 to move it to page 6-7.

• A new Containment Inspection Report is being added as TS 6.6.6 which states:

Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Tendon Surveillance Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

These reporting requirements are identical to those previously approved by the NRC in the ANO-1 Technical Specifications TS 5.6.6, Reactor Building Inspection Report.

The guidance in NUREG-1432, Revision 3, is enveloped in the ANO-1 TSs and the proposed ANO-2 TSs.

• The revised Bases to TS 4.6.1.5 are being modified to be consistent with NUREG-1432, Revision 3 except for the reference to RG 1.35. These Bases are being provided for information only.

3.0 BACKGROUND

10 CFR 50.55a(b)(2) requires licensees to implement the requirements of Subsection IWL of the ASME Code. The notice in the Federal Register (61 FR 413030) recognized that the final rule had satisfactorily considered the previous guidance provided in Regulatory Guide 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Revision 3.

ANO has established a Containment Inspection Program which includes both the tendon surveillance inspection requirements and the containment surface inspection requirements as required by Subsection IWL and 10 CFR 50.55a.

The containment building structure is discussed in the ANO-2 Safety Analysis Report (SAR)

Section 3.8.1.1. The containment consists of three basic parts: (1) a flat circular base slab, (2) a right circular cylinder, and (3) a sphere-torus dome. The containment is constructed of reinforced concrete prestressed by post-tensioned tendons in the cylinder and the dome.

Special reinforcing details are provided at discontinuities. The design of the containment is described more specifically in SAR Section 3.8.1.4.

The cylinder wall is prestressed by a system of horizontal and vertical tendons. The horizontal tendons are anchored at three buttresses equally spaced around the outside of the containment. The vertical tendons are anchored to the base slab at bottom and the ring girder at top. The dome is prestressed by three systems of dome tendons spaced at 120 degrees apart. The three-way dome tendons are anchored at the side of the ring girder. The tendons are installed in sheaths which are filled with a corrosion inhibitor. An access gallery is provided beneath the base slab for installation of the vertical tendons and inspection of bottom anchorage.

to 2CAN1 20401 Page 3 of 6 The interior of the containment is lined with steel plates welded together to form a leak tight barrier. Since the base slab liner plate is covered with concrete, leak chase channels are provided at seam welds, to allow for leak testing during normal operation.

4.0 TECHNICAL ANALYSIS

4.1 Removal of Existing ANO-2 TS 4.6.1.5 Surveillance Requirements The current Surveillance Requirement (SR) 4.6.1.5.1, Containment Tendons requires the visual examination of twenty-one tendons per specified interval. In addition, the SR requires the satisfactory completion of the examinations performed at the 1, 3, and 5 year interval and subsequent 5 year intervals. The requirements of Surveillance 4.6.1.5.1 have been superseded by the requirements contained in Subsection IWL and 10 CFR 50.55a and have therefore been deleted.

The requirements of Specification 4.6.1.5.2 have been superseded by the requirements contained in Subsection IWL of the ASME Code, 10 CFR 50.55a, and have therefore been deleted.

The requirements of Surveillance 4.6.1.5.3 are redundant to the requirements for a Containment Inspection as contained in 10 CFR 50, Appendix J. ANO-2 TS Section already requires compliance with the Containment Leakage Rate Testing Program contained in TS 6.15. Therefore, this SR is not required.

4.2 Containment Tendon Surveillance Program The portion of the ANO Containment Inservice Inspection Program which ensures the structural integrity of the containment through inspection of the tendon system is the ANO Containment Tendon Surveillance Program. Details of ANO Containment Inservice Inspection Program are provided in Entergy Procedure CEP-CI1-007, Arkansas Nuclear One Units I and 2 Containment Inservice Inspection (CII) Program Plan. Compliance with TS 6.5.6 will be performed by application of this program.

The Containment Tendon Surveillance Program, inspection frequencies, and acceptance criteria are currently in accordance with the 1992 Edition, 1992 Addenda of ASME Code, Section Xl, Subsection IWL and 10 CFR 50.55a. Entergy will update to later ASME Code Editions as required by 10 CFR 50.55a. However, Revision 3 of NUREG-1432 refers to a bracketed application of Regulatory Guide 1.35, Revision 1 for implementing a Containment Tendon Surveillance Program. Instead, Entergy is implementing the Containment Tendon Surveillance Program as an integral part of the Containment Inservice Inspection Program. This program is in compliance with the requirements of 10 CFR 50.55a(b)(2). The previous Containment Tendon Surveillance Program based on Regulatory Guide 1.35, Revision 1, has been superseded based on the issuance of Subsection IWL of the ASME Code.

Upon identification of any degradation reaching specific thresholds defined by Subsection IWL of the ASME Code, an Engineering Evaluation is required to determine the impact of the degradation on overall operability. If structural integrity cannot be established from the Engineering Evaluation, the Containment will be declared inoperable. By including to 2CAN120401 Page 4 of 6 structural integrity in Specification 3.6.1.5, the unit would be required to restore structural integrity (Operability) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or commence plant shutdown. Specification 3.6.1.5 currently allows a restoration period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before unit shutdown must commence.

The proposed change is, therefore, more restrictive and is generally consistent with that of NUREG-1432.

Containment Inspection Program Reporting The reporting requirements under TS 6.6.6 will require an Engineering Evaluation be submitted to the NRC within 30 days of completion for any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Tendon Surveillance Program. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

This reporting requirement establishes the timing for submittal of the information required by IWL-331 0. This reporting requirement is also consistent with that contained in the ANO-1 Technical Specification 5.6.6 issued in ANO-1 License Amendment 199 (Ref. 1).

Even though the wording is the same as the current ANO-1 TSs, it includes the considerations of NUREG 1432.

Based on the above, Subsection IWL of the ASME Code and 10 CFR 50.55a(b)(2) adequately address testing of the containment structure. Incorporating current requirements, the elimination of redundant regulations, and implementing administrative improvements provide technical specifications that are more appropriate. Because existing requirements are controlled by regulation, there is no reduction in safety and adequate control is maintained. ANO-2 has determined that the controls governing the concrete containment inspection provided by the ANO Containment Inservice Inspection Program are adequate and are in keeping with the philosophy associated with the NUREG-1432, Revision 3.

5.0 REGULATORY ANALYSIS

5.1 Safety Analysis Report (SAR) Review ANO-2 SAR Section 3.8.1.7.3 Inservice Tendon Surveillance states:

The objective of the inservice tendon surveillance program during the lifetime of the plant is to provide a systematic means of assessing the continued quality of the post-tensioning system. The program is intended to furnish sufficient inservice historical evidence to provide a measure of confidence in the condition and the functional capability of the system, as well as an opportunity for timely corrective measures should adverse conditions, such as excessive corrosion, be detected. The inservice tendon surveillance program will be conducted in accordance with the requirements of ASME B&PV Code Section Xl Subsection IWL as modified by 10 CFR 50.55a.

This statement represents required compliance with ASME B&PV Code Section Xl Subsection IWL and is consistent with the proposed changes in Attachment 2 to this submittal.

i.

Attachment 1 to 2CAN120401 Page 5 of 6 General Design Criterion 53, Provisions for Containment Testing and Inspection, requires that the reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations which have resilient seals and expansion bellows. Entergy assures that the ANO-2 containment retains its ability to retain design basis pressure by the application of the ANO Containment Tendon Surveillance Program being required by 10 CFR 50.55a and the proposed TS.

5.2 Determination of No Significant Hazards Consideration Entergy Operations, Inc. is proposing that the Arkansas Nuclear One Unit 2 (ANO-2)

Operating License be amended to revise the requirements for ensuring containment structural integrity. The proposed changes modify the Containment Structural Integrity Technical Specification (TS) 3.6.1.5 to delete the existing Surveillance Requirements (SR) and add a new SR to verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. A new Containment Tendon Surveillance Program is added to TS 6.5.6 and a new reporting requirement is being added to TS 6.6.6. The proposed changes are generally consistent with NUREG 1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.

An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The containment building is not considered to be the initiator of any accident previously evaluated, but serves to mitigate accidents that could allow a release to the environment. The proposed TS change will provide for containment tendon inspections as required by 10 CFR 50.55a and prevent or inhibit release from the containment building as designed. Through appropriate inspections and implementation of corrective actions for any degradation discovered during the inspections that might lead to containment structural failures, the probability or consequences of accidents will not be increased.

Therefore, the removal of inspection details from the TS does not involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed change does not change the design, configuration, or method of operation of the plant. By implementing corrective actions for any degradation discovered during the required inspections of the containment, the possibility of a new or different kind of accident will not be created. Implementation of the requirements of Subsection IWL of the ASME code and those of 10 CFR 50.55a(b)(2) provide an equally acceptable containment inspection program.

to 2CAN120401 Page 6 of 6 Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed change to incorporate the applicable requirements of Subsection IWL of the ASME Code and of 10 CFR 50.55a(b)(2) into the ANO-2 containment inspection program has no impact on any safety analysis assumptions. The addition of structural integrity requirements to ANO-2 TS Specification 3.6.1.5 imposes consistent requirements with those previously specified in the ANO-2 TSs. The requirements of ASME IWL are more restrictive than those currently provided in the existing ANO-2 technical specifications. As a result, the margin of safety is not reduced by the proposed change.

Therefore, this change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does not involve a significant hazards consideration.

5.3 Environmental Considerations The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to record keeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE The above proposed change for the addition of the reporting requirement is similar to that issued for ANO-1 in Reference 1. The change to meet the current revised standard Technical Specifications is similar to that approved for South Texas Project in the NRC Safety Evaluation dated March 19, 2002 and for the Virgil C. Summer Nuclear Station in Safety Evaluation dated September 6, 2000.

Attachment 2 2CAN1 20401 Proposed Technical Specification Changes (mark-up)

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.50PERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

If the containment is not OPERABLEWith the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits within 24-1 hours 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 />.

SURVEILLANCE REQUIREMENTS 4.6.1.5.1 Containment Tcndons The containment tendons' structural integrity shall be demonstrated at the end of one, three and five years following the initial cRntainment structural integrity test and at five year intervals thereafter. The tendons' structural integrity shall be demonstrated by a vih/a! earnination (to the extent practical and without dismantling load bearing components of he anchoFage) of a representative sample* of at least 21 tendons (6 dome, 5 veFtiial, and 10 hoop) and vethyng-.no abnormal-degradation. Unless there is evidenGe of abnornal degradation of the containment tendons during the first three tests of the tendons, the number-o tendons examined during subsequent tests may be reduced to a representative sample of least 9 tendons (3 dome, 3 vertical and 3 hoop).

Verify containment structural integrity in accordance with the Containment Tendon Surveillance Proqram.

• For each inspection, the tendons shall be selected on a random but-repfesentative basis-so thatahe sample group will change somewhat for each inspection; however, to develoa history of tendon performance and to correlate the observed data, one tendon from each group (dome, v-rtisal,-and hoop) may be kept unchanged after heinia6eletin ARKANSAS - UNIT 2 3/4 6-8 Amendment No.

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I kflrlRITI-t rn~lrTWrlnIl CfnQ n~D:DAT-flnI 4.6.1.5.2 End Anchoraqes and Adjacent Concrete Surfaces The 6tructural integFity ef the cnd anchorages of all tendons inspected pursuant to Specification 4.6.1.5.1 and-the nt conct urfars rhall be demonstrated by determining throughinspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjac;nt to the enrd anhorages.

Inspections of the concrete shall be performed during the Type A containment leakage rate testE (reference Specification 4.6.1.2) while the containment is at its maiumbtest suFe.

4..6.4.5.3 Gontainment Suerfaces :The stFUctural ntegity of the exposed acressible interioF and exterior surfaces of the containment, including the liner plate, shall be determiRned by a visual inspection f these surfaces and verifying e appareRnt changes in appearance or other abnormal degradation has occurred in accordance with the Containment Leakage Rate TeSting Program.

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I' CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.6 The containment purge supply and exhaust isolation valves shall be closed and handswitch keys removed.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one or more containment purge supply and/or exhaust isolation valves not closed with the handswitch keys removed, place the valve(s) in the closed position with handswitch keys(s) removed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 />.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The containment purge supply and exhaust isolation valves shall be determined closed at least once per 31 days.

ARKANSAS - UNIT 2 3/4 6-9a Amendment No. 64 l

ADMINISTRATIVE CONTROLS 6.5.4 Radioactive Effluent Controls Program (continued)

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix l;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l;

i. Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from iodine-131, iodine-1 33, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l; and

j. Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

6.5.5 Component Cyclic or Transient Limit Program This program provides controls to track the SAR Section 5.2.1.5, cyclic and transient occurrences to ensure that components are maintained within the design limits.

6.5.6 not usedContainment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The Containment Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with the ASME Code, Section Xl, Subsection IWL and 10 CFR 50.55a.

The provisions of SR 4.0.3 are applicable to the Containment Tendon Surveillance Program inspection frequencies.

ARKANSAS - UNIT 2 6-6 Amendment No.2-%

6.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory Position C.4.b of Regulatory Guide 1.14.

Revision 1, August 1975. The volumetric examination per Regulatory Position C.4.b.1 will be performed on approximately 10-year intervals.

6.5.8 Inservice Testina Proaram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies specified in Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Code Required frequencies terminology for for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Every 6 weeks At least once per 42 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice testing activities.

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

ARKANSAS - UNIT 2 6-7 Amendment No. 255

ADMINISTRATIVE CONTROLS 6.6.6 not-used Containment Inspection Report Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Tendon Surveillance Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

6.6.7 Steam Generator Tube Surveillance Reports

a. Following each inservice inspection of steam generator tubes the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

b. The complete results of the steam generator tube inservice inspection shall be reported within 12 months following the completion of the inservice inspection.

This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission as denoted by Table 6.5.9-2. Notification of 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 determine cause of the tube degradation and corrective measures taken to prevent recurrence.

6.6.8. Specific Activity 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 6-23 Amendment No. 255

Attachment 3 2CAN1 20401 Proposed Technical Specification Bases Changes (mark-up)

CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE AND AIR TEMPERATURE The limitations on containment internal pressure and average air temperature, assuming a worst case relative humidity value of 0 %, ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 5.0 psi, 2) the containment peak pressure does not exceed the design pressure of 59 psig during design basis conditions, 3) the ECCS analysis assumptions are maintained, and 4) the containment cooling fan motor qualifications are maintained.

The limitation on containment average air temperature ensures that the containment liner plate temperature does not exceed the design temperature of 300OF during LOCA conditions. The containment temperature limit is consistent with the accident analyses. Figure 3.6-1 represents analysis limits and does not account for instrument error.

3/4.6.1.5 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum design pressure of 59 psig.

Thevisual exaHiRatnRofetepons, aRnhorages and Gcntainmernt uares and- the Type-A leakage tests of the Unit 2 containment in conjunction with-he required surveillance activities of the-UnitJ 1on-taiment arc sufficient to demonstrate this Gapabily For ungrouted, post tensioned tendons, the SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program. Testing and frequency are consistent with 10 CFR 50.55a(b)(2). and Subsection IWL of the ASME Code.

The swupeillan uirments for demonstrating the containment's structural integrity arc in Gcmpliance with the recommendations of Regulatory Guide 1.35 "Inservice Surveillance of U1ngruted Tendons in Proetressed Concrete ^^ntainmet-Stfulturesa u 76 3/4.6.1.6 CONTAINMENT VENTILATION SYSTEM The containment purge supply and exhaust isolation valves are required to be closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system.

ARKANSAS - UNIT 2 B 3/4 6-2 Amendment No. 439,-76,225