1CAN049901, Application for Amend to License DPR-51,revising Requirements Associated with ANO 1 Provisions for RB Testing & Insp

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Application for Amend to License DPR-51,revising Requirements Associated with ANO 1 Provisions for RB Testing & Insp
ML20205P386
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 04/09/1999
From: Hutchinson C
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205P389 List:
References
1CAN049901, 1CAN49901, NUDOCS 9904200161
Download: ML20205P386 (13)
v • d • e


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-;g- Entergy Operations,Inc.

C. Randy Hutchinson

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a. . .m April 9,1999 1CAN049901 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station OPI-17 '

Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 1 Proposed Technical Specification Changes Revising Reactor Building Structural l Integrity Requirements l Gentlemen:

l Attached for your review and approval are proposed Technical Specification (TS) changes revising the requirements associated with Arkansas Nuclear One - Unit 1 (ANO-1) provisions for reactor building testing and inspection. The proposed changes affect ANO-1 TS Limiting Conditions for Operation (LCO), Surveillance Requirements, and applicable Bases relevant to

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/j inservice inspection requirements for the containment structures, tendons, and anchorages.

The proposed changes are a result of the revisions to 10 CFR 50.55a published in the Federal Register (61 FR 41303), made effective on September 9,1996, and required to be fully flj iniplemented by September 9, 2001. These revised requirements affect the surveillance j methods for the containment tendons, the conduct of containment visual inspections, and the reporting methods employed in disseminating the results of these inspections to the NRC.

ANO-1 is implementing a containment inspection program to comply with the revised requirements of 10 CFR 50.55a. The program is based on the American Society of l Mechanical Engineers (ASME)Section XI, Subsection IWL, as required and modified by l revised 10 CFR 50.55a. The proposed change is consistent with the current regulatory requirements as described in 10 CFR 50.55a(g)(6)(ii)(B). Therefore, it does not result in the l reduction of previous TS commitments to implement an effective containment surveillance l program.

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 the attached submittal. '

9904200161 990409 l PDR ADOCK 05000313 p PDR

U. S. NRC April 9,1999 l ICAN049901 Page 2 The proposed changes are intended to eliminate regulatory redundancy and historical testing requirements. Since surveillance of the ANO-1 reactor building will commence prior to unit  !

shutdown for refueling outage IR15, currently scheduled to commence on September 10, 1999, Entergy Operations requests NRC approval by August 1,1999, with an implementation period of 30 days.

Very truly you , j 7/

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Attachment To the best of my knowledge and belief, the statements contained in this submittal are true. I SUBSCRIBED AND SWORN TO before me, a Nota Public in and for  %

County and the State of Arkansas, this 9" day of Od ,1999. '

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" OFFICIAL SEAL"  ?

Andrea Pierce u, O_-- y q Notary Public, State of Arkanm ',

Notary Public LM My Commission Expires /A//5/4oo 7 "==ycomm$UonNp3.jiscoof

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U. S. NRC April 9,1999 1CAN049901 Page 3 cc; Mr. Ellis W. Merschoff Regional Administrator )

U. S. Nuclear Regulatory Commission

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Region IV 611 Ryan Plaza Drive, Suite 400

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Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 l l

Mr. Nick Hilton NRR Project Manager Region IV/ANO-1 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-D-18 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. David D. Snellings Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205  !

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ATTACitMETjI IQ lCAN049901 PROPOSED TECHNICAL SPECIFICATION AND RESPECTIVE SAFETY ANALYSES IN TIE MATTER OF AMENDING LICENSE NO. DPR-51 1

ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT ONE DOCKET NO. 50-313 l

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i Attachment to ICAN049901 Page1of9 l

l DESCRIl9' ION OF PROPOSED CHANGES The proposed changes to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) are as follows:

  • The TS Table of Contents and List of Figures have been updated to reflect the changes made to ANO-1 specifications relevant to this submittal.

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  • Terminology used in Specification 3.6.1, its action statement, and the Objective of l Specification 3.6 have been revised to permit the inclusion of additional requirements l beyond those defined in Specification 1.7, " Reactor Building."
  • The bases for Specification 3.6 have been revised to clarify the reactor building operability requirements. In addition, applicable code requirements are also listed for future reference, where applicable.
  • The footer on Page 80 of the ANO-1 TS has been modified to inform the user that several of the following pages have been deleted (some due to this submittal while others had previously been deleted). No technical changes were made to this page.
  • The specific surveillance criteria of Specification 4.4.2 is programmatic in nature and has therefore been deleted from the specifications and incorporated in ANO-l's containment i inspection program.
  • Figures 4.4.2-1, 4.4.2-2, and 4.4.2-3 are also programmatic in nature and have been l

deleted from the specifications and relocated to ANO-l's containment inspection program.

  • The requirement of Specification 6.12.5.a has been deleted. Reporting requirements l

associated with this program are incorporated in ANO-l's containment inspection program.

BACKGROUND i

The reactor building structure is discussed in the ANO-1 Safety Analysis Report (SAR) i Section 5.2 as a reinforced prestressed concrete structure in the shape of a cylinder with a shallow domed roof and a flat foundation slab.

The cylindrical portion is prestressed by a post-tensioning system consisting of horizontal and vertical tendons. The dome has a 3-way post-tensioning system. Hoop tendons are placed in 3-240 degree systems using three buttresses as anchorages. The foundation slab is conventionally reinforced with high strength reinforcing steel and is founded on bedrock. A continuous access gallery is provided beneath the base slab for installation of vertical tendons.

, s Attachment to ICAN049901 Page 2 of 9 A welded steel liner is attached to the inside face of the concrete shell to insure a high degree ofleak tightness. The base liner is installed on top of the structural slab and is covered with concrete. The structure provides shielding for both normal and accident conditions.

On January 7,1994, the NRC published in the Federal Register (59 FR 979) a proposed amendment to its regulation,10 CFR 50, " Domestic Licensing of Production and Utilization Facilities," to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE, and Subsection IWL, of Section XI, Division 1, of the AShE Code. Following the required comment period, the NRC issued the final rule effective on September 9,1996. The final rule,10 CFR 50.55a(g)(6)(ii)(B), required licensees to implement the requirements of Subsection IWE and Subsection IWL of the AShE Code, with specified modifications and limitations, by September 9, 2001. The notice in the Federal Register (61 FR 413030) recognized that the fmal 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-1 is implementing a containment inspection program in compliance with the requirements of 10 CFR 50.55a(g)(6)(ii)(B). Therefore, the previous commitment to a containment inspection program based on Regulatory Guide 1.35, Revision 2, has been superseded by the final rulemaking, and the proposed change is necessary to improve the ANO-1 TS to conform to these new regulatory requirements. Note that the ANO-1 program incorporating Subsection IWE of the ASME code will be completed in the near future and is not included with this submittal. Existing requirements for those components associated with Subsection IWE of the AShE code will remain intact until a revised program is implemented.

Current surveillance requirements associated with the centainment inspection program located in the ANO-1 TS were drafted from guidelines provided under Regulatory Guide 1.35, Revision 2. In light of the regulations established in 10 CFR 50.55a(g)(6)(ii)(B) and subsequent detail provided in Subsection IWL of the AShE code, revising the subject surveillance requirements from the ANO-1 TS is necessary. In addition, several existing requirements in the ANO-1 TS relate solely to initial testing and have since been satisfied.

ANO has developed a containment inspection program consistent with the new regulations relevant to Subsection IWL, detailed in draft procedure 5220.011,"ANO 1 & 2 Containment Building Tendon Surveillance and Concrete Inspection." In progressing toward the Improved Standard Technical Specification (ITS) format, ANO-1 proposes to remove the level of detail currently located in its TS's to the ANO-1 containment inspection program. The aforementioned procedure is safety related and therefore requires multiple levels of site approval and review prior to changes being implemented. Sufficient controls exist under the procedure change process at ANO-1 to ensure current and future regulations and commitments are properly addressed when making revisions to the containment inspection procedure. In addition, the current condition reporting program at ANO is suflicient to ensure appropriate degradation of any containment structure or component associated with procedure 5220.011 is properly evaluated and, if appropriate, reported under requirements of

, o, Attachment to 1CAN049901 Page 3 of 9 10 CFR 50.72 and 50.73. Based on the above, ANO-1 has determined that the controls governing the concrete containment inspection provided by the ANO-1 containment inspection program are adequate without additional TS controls and are in keeping with the philosophy associated with the ITS.

Similar submittals were made by Florida Power Corporation (FPC) representing the Crystal River Unit 3 facility dated February 19, 1999, and by Baltimore Gas and Electric Company (BGE) representing the Calvert Cliffs Unit I and Unit 2 facilities dated November 20,1998.

Both FPC and BGE have similar reactor building structures as that of ANO-1 and the submittals revised administrative, reportability, and surveillance sequirements from the TSs to I an appropriate containment inspection program. The customized ANO-1 TSs require changes l in addition to those submitted by FPC and BGE in order to incorporate the philosophy in the .

ITS. Specifically, the ANO-1 Limiting Conditions for Operations (LCO) associated with reactor building integrity is revised to be consistent with ITS. Also, unlike FPC and BGE, ANO-1 is submitting changes relevant to Subsection IWL requirements only; IWE requirements will be addressed at a later date.

DISCUSSION OF CIIANGE 1

Several sections of the ANO-1 TSs are affected by this submittal. In addition, changes range j from those requiring some detailed discussion to those that are solely administrative in nature. .

As a result, most of the affected TSs will be discussed individually ta promote increased clarity and readability. Changes to Specification 4.4.2 require greater detail and therefore will be discussed at length. Each subsection under the Specification 4.4.2 is discussed separately below.

4.4.2.1 Tendon Surveillance The first paragraph of this specification requires the testing of twenty-one tendons per specified interval. However, as stated in the second paragraph, satisfactory completion of the tests performed at the 1, 3, and 5 year interval allows decreasing the required number of  ;

tendons tested per interval to nine. Transmittals previously addressed to the NRC, I ICAN097508, dated September 11, 1975, ICANil7704, dated November 4,1977, and ICAN087909, dated August 22,1979, stated the satisfactory completion of this requirement and therefore only the second paragraph remains in effect. The second paragraph of Specification 4.4.2.1 is being modified to comply with the requirements of the ASME Code j for Class CC components based on Subsection IWL of the ASME Code and 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). These requirements are being incorporated into the ANO-1 containment inspection program. Therefore, the requirements of Specification 4.4.2.1 are redundant to the requirements contained in Subsection IWL and have been deleted.

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0 Attachment to 1CAN049901 Page 4 of 9 4.4.2.1.1 Lift Off The requirements of Specification 4.4.2.1.1 are being modified to comply with the requirements of the ASME Code for Class CC components based on Subsection IWL of the ASME Code and 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). These requirements are being incorporated into the ANO-1 containment inspection program. Therefore the requirements of Specification 4.4.2.1.1 are redundant to the requirements contained in Subsection IWL and have been deleted.

4.4.2.1.2 Wire Inspection and Testing The requirements of Specification 4.4.2.1.2 are being modified to comply with the requirements of the ASME Code for Class CC components based on Subsection IWL of the ASME Code and 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). These requirements are being incorporated into the ANO-1 containment inspection program. Therefore the requirements of Specification 4.4.2.1 are redundant to the requirements contained in Subsection IWL and have been deleted. In addition, the last sentence of this specification is largely redundant to Specification 4.4.2.1.3(3).

4.4.2.1.3 Acceptance Criteria The contents of this specification contain general criteria only, listing little if any specific values in which comparents are to be tested. This makes the specification somewhat ambiguous in nature. The requirements of ASME Code for Class CC components based on Subsection IWL of the ASME Code and 10 CFR 50.55a(g)(6)(ii)(B) encompass the intent of this specification and in greater detail, providing absolute values in which comparisons of tested components shall be made. For example, the values listed in Part 3 of this specification are not only included in the program, but an additional requirement to determine the base neutralization number of the sheathing filler material is rJso incorporated from the code.

Corresponding Figures 4.4.2-1, 4.4.2-2, and 4.4.2-3 are incorporated directly into the program being developed. ANO-1 has determined that the containment inspection program properly incorporates the requirements of Subsection IWL and 10 CFR 50.55a(g)(6)(ii)(B) and therefore the deletion of this specification along with the three aforementioned TS Figures is acceptable.

4.4.2.2 Insoection Intervals and Reports The required inspection intervals stated in this specification are consistent with IWL-2400, Inservice Inspection Schedule of Subsection IWL of the ASME code and are therefore being incorporated into the ANO-1 containment inspection program. Reportability requirements are more specifically addressed under 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) and, IWA-6000 of Subsection IWA and IWL-3310 of Subsection IWL, Evaluation Report, of the ASME code and will be incorporated into the ANO-1 containment inspection program as set forth in the code. The requirements of Specification 4.4.2.1 are redundant to the requirements contained in Subsection IWL and Subsection IWA and have therefore been deleted.

Attachment to 1CAN049901 Page 5 of 9 4.4.2.3 End Anchorage Concrete Surveillance Requirements associated with Specification 4.4.2.3, including the stated intervals for testing, have been successfully completed and do not contain any future testing requirements. ANO letter ICAN037504, dated March 24,1975, communicated the successful completion of the testing requirements in this specification. The requirements associated with this specification are historical in nature and no longer pertain to current testing regulations. As a result, this change is considered administrative in nature. Therefore, since the requirements of this  ;

specification have been successfully completed and this specification contains no future testing l requirements, this specification has been deleted.

4.4.2.4 Liner Plate Surveillance Requirements associated with Specification 4.4.2.4 including the stated intervals for testing have been successfully completed. Specification 4.4.2.4.4 allowed this surveillance program to be discontinued after the inspection made during the refueling shutdown if no corrective action was needed. Satisfactory inspection results were documented under the Arkansas Power and Light Company (AP&L) ANO-1 Reactor Building Structural Integrity Test '

Report, dated January 18, 1974, and under ANO procedure 1304.086 Revision 0, Reactor i Building Liner Plate Surveillance Procedure, dated February 24, 1977. The requirements l associated with this specification are historical in nature and no longer pertain to current I testing regulations. As a result, this change is considered administrative in nature. Therefore, since the requirements of this specification have been succes.sfully completed and this specification contains no future testing requirements, this specification has been deleted. l l

BASES Structural Integrity l

Given the relocation, or where applicable, deletion of surveillance requirements relevant to reactor building structural integrity, the bases for this section have also been deleted.

Applicable useful statements concerning the general bases of maintaining the structural integrity of the reactor building operable have been relocated or added to the bases associated with Specification 3.6.1 (discussed later in this section).

Although a specific LCO currently does not exist in the ANO-1 TS addressing reactor building structural integrity, ANO-1 believes it is prudent to ensure the operator or user is aware of requirements associated with the operability of the reactor building. The existing Specification 3.6.1 sets forth requirements for maintaining the integrity of the reactor building.

Reactor building stmetural integrity is not specifically addressed in the definition of reactor building integrity of Specification 1.7. Therefore, to ensure the user is aware of requirements associated with structural integrity, the terminology of Specification 3.6.1, its associated action statement, and its associated bases has been changed to reflect these requirements.

Operability, as defined in Specification 1.3, sufficiently covers any part of a structure or component required to be operable in order to meet specific safety functions. By changing the terminology of Specification 3.6.1 from maintaining the integrity of the reactor building to

b Attachment to \

ICAN049901 Page 6 of 9 maintaining the reactor building operabk, the requirements of stmetural integrity may be included. The bases of Specification 3.6.1 provides an explanation to the user, discussing that operability in this case includes both reactor building integrity as defined in Specification 1.7 and reactor building structural integrity as provided by Subsection IWL of the ASME code and 10 CFR 50.55a(g)(6)(ii)(B). By including structural integrity in Specification 3.6.1, the l unit would be required to restore structural integrity (operability) within I hour or commence

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plant shutdown. Upon identification of any degradation reaching specific thresholds defined i by Subsection IWL of the ASME code, an engineering evaluation is required to determine the  ;

impact of the degradation on overall operability. This action is consistent with the current requirements of ANO-1 TSs and the ITS.

The requirements ofIWL-3300 of Subsection IWL and IWA-6000 of Subsection IWA of the ASME code, and of 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) adequately address reportability of containment inspections performed. Specific failures identified during testing are evaluated under the licensee's condition reporting program and a reportability determination is completed pursuant to 10 CFR 50.72 and 10 CFR 50.73. Therefore, the i special report requirement of Specification 6.12.5(a), Tendon Surveillance, Specification  !

4.4.2.2, has been deleted.

Administrative changes were made to the ANO-1 TS Table of Contents and List of Figures, removing those statements referring to specifications deleted as discussed previously in this submittal. In summation, Subsection IWL of the ASME code,10 CFR 50.55a(g)(6)(ii)(B),

10 CFR 50.72, and 10 CFR 50.73 adequately address testing and reportability of the containment structure and therefore, ANO-1 requests the aforementioned changes to its technical specifications be approved. Sufficient controls exist under the procedure change process at ANO-1 to ensure current and future regulations and commitments are properly addressed when making revisions to the containment inspection procedure. Incorporating current requirements, the elimination of redundant regulations, and implementing administrative improvements provide technical specifications that are less cumbersome and user friendly. Because existing requirements are controlled by regulation, there is no reduction in commitment and adequate control is maintained.

DETERMINATION OF NO SIGNIFICANT IIAZARDS CONSIDERATION Entergy Operations, Inc. is proposing that the Arkansas Nuclear One Unit 1 (ANO-1)

Operating License be amended to revise the requirements associated with the inspection and testing of the containment building structure. The current requirements for the containment inspection program based on Regulatory Guide 1.35, Revision 2, have been superseded by the revised regulations. The proposed change is necessary to update the ANO-1 TS to conform to these new regulatory requirements. The proposed changes are consistent with the requirements of ASME Code for Class CC components based on Subsection IWL of the ASME Code and 10 CFR 50.55a(g)(6)(ii)(B).

, I Attachment to 1CAN049901 Page 7 of 9 An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(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 proposed change to the ANO-1 TS replaces previous requirements and commitments to establish a containment inspection program based on the guidance provided in Regulatory Guide 1.35, Revision 2 in favor of regulations depicted in 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix). ANO-1 is implementing a containment inspection program to comply with these new regulatory requirements.

The final rule specifies requirements to assure that the critical areas of the containment stmeture are routinely inspected to detect and take corrective action for defects that could compromise structural integrity.

Maintaining reactor building structural integrity is independent of the operation of the ,

reactor coolant system (RCS), the reactor protection system (RPS) and emergency I core cooling system (ECCS). The reactor building is not considered to be the initiator of any accident previously evaluated. The physicallocation ofinspection details dcses not prevent or inhibit the reactor building from functioning as designed to provide an acceptable barrier against release of radioactive materials to the environment.

Through appropriate inspections and implementation of corrective actions for any l 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 ofinspection details from the TS does aqt 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.

Maintaining containment structural integrity is independent of the operation of the RCS, the RPS and ECCS. The proposed changes do 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(g)(6)(ii)(B) and 50.55a(b)(2)(ix) provide an equally acceptable containment inspection program.

Therefore, this change does nqt create the possibility of a new or different kind of accident from any previously evaluated.

X Attachment to 1CAN049901 Page 8 of 9 l

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

The removal of the level of detail currently found in the ANO-1 TS regarding reactor building inspections and incorporating the applicable requirements of Subsection IWL l of the ASME code and of 10 CFR 50.55a(g)(6)(ii)(B) and 50.55a(b)(2)(ix) into the l ANO-1 containment inspection program has no impact on any safety analysis assumptions. Requirements associated with containment inspections are controlled by safety related procedure 5220.011. Suflicient controls exist under the procedure change process at ANO-1 to ensu:e current and future regulations and commitments are properly addressed when making revisions to the containment inspection procedure. The addition of structural integrity requirements to ANO-1 TS Specification 3.6.1 imposes consistent requirements with those previously specified in the ANO-1 TSs. The containment inspection program ensures that the containment will function as designed to provide an acceptable barrier against release of radioactive materials to the environment. Through the implementation of the containment inspection program, the existing margin of safety is preserved.

Therefore, this change does ag! involve a significant reduction in the margin of safety.

Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does nR1 involve a significant hazards consideration.

ENVIRONMENTAL IMPACT EVALUATION 10 CFR 51.22(c) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released off-site, or (3) result in a significant increase in individual or cumulative occupational radiation exposure. Entergy Operations, Inc. has reviewed this license amendment and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the isseance of the proposed license amendment. The bases for this determination is as follows:

1. The proposed license amendment does not involve a significant hazards consideration as described previously in the evaluation.

Attachment to 1CAN049901 Page 9 of 9

2. As discussed in the significant hazards evaluation, this change does not result in a significant change or significant increase in the radiological doses for any Design Based Accident. The proposed license amendment does not result in a significant change in the types or a significant increase in the amounts of any effluents that may be released off site.
3. The proposed license amendment does not result in a significant increase to the individual or cumulative occupational radiation exposure because this does not modify the method of operation of systems and components necessary to prevent a radioactive release.
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