2CAN060403, Unit 2 - Request for Additional Information Responses for License Renewal Application TAC No. MB8402

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Unit 2 - Request for Additional Information Responses for License Renewal Application TAC No. MB8402
ML041700183
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 06/16/2004
From: Mitchell T
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
-RFPFR, 2CAN060403, TAC MB8402
Download: ML041700183 (20)
v • d • e


Text

Entergy Entergy Operations, Inc.1448 S.R. 333 Russellville, AR 72802 Tel 501 858 5000 2CAN060403 June 16, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Request for Additional Information Responses for License Renewal Application TAC No. MB8402 Arkansas Nuclear One -Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

By letter dated May 17, 2004 (2CNA050404), the NRC requested additional information on the Arkansas Nuclear One, Unit 2 (ANO-2) License Renewal Application (LRA) within 30 days of receipt. The requests for additional information (RAls) are from the LRA Section 4.3, Metal Fatigue, Section 4.5, Concrete Containment Tendon Prestress, Section 4.6 Containment Liner Plate and Penetration Fatigue Analyses, and Section 4.7.6 High Energy Line Break. The responses to the RAls are contained in Attachment 1.The new commitment contained in this submittal is summarized in Attachment

2. Should you have any questions concerning this submittal, please contact Ms. Natalie Mosher at (479) 858-4635.I declare under penalty of perjury that the foregoing is true and correct. Executed on June 16, 2004.Tigothy G. Mitchell Director, Nuclear Safety Assurance TGM/nbm Attachments A iuaD 2CAN060403 Page 2 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 Mail Stop 0-11 F1 Washington, DC 20555-0001 U. S. Nuclear Regulatory Commission Attn: Mr. Greg Suber Mail Stop 0-11 F1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Attachment I 2CAN060403 RAI Responses Attachment 1 to 2CAN060403 Page 1 of 15 Section 4.3, 4.5, 4.6, and 4.7.6 RAI Reponses RAI 4.3.1-1: Provide the edition of the American Society of Mechanical Engineers (ASME)Section III Code that is applicable to the Class 1 fatigue analysis, and indicate if it was reconciled with the initial edition used for construction.

Response:

The reactor coolant system (RCS) mechanical components evaluated for fatigue include Combustion Engineering-designed vessels (e.g., reactor vessel and control element drive mechanism (CEDM) pressure boundary, and pressurizer), reactor coolant pumps, RCS piping, and the Westinghouse replacement steam generators.

These components were designed in accordance with the following editions of ASME Section III: Component Applicable Code Year-ASME IlIl* Reactor Vessel and Pressurizer 1968 Edition through Summer 1970 Addenda.CEDMs 1971 Edition through Winter 1972 Addenda.RCS Piping 1971 Edition through Summer 1971 Addenda and 1980 Edition.Replacement Steam Generators 1989 Edition, no Addenda* Reactor Coolant Pumps 1971 Edition Construction is understood to include material selection, design, fabrication, examination, testing, overpressure relief, marking, stamping, and preparation of applicable reports as required by the design code of record. For the reactor vessel, pressurizer, and CEDM the design codes listed above and construction codes are equivalent.

RCS piping was constructed in accordance with ASME Section 111, 1971 Edition through Summer 1971 Addenda. Fatigue evaluations for selected Class 1 piping'were revised in accordance with ASME Section 1II, 1980 Edition to support the steam generator replacement project.Specifically, Class 1 piping stress reports were revised to qualify piping lines for the revised seismic spectra and anchor movements for the reactor building and reactor coolant system.The replacement steam generators were constructed in accordance with ASME 111, 1989 Edition. No reconciliation is required since the edition of the Code that is applicable to the fatigue analysis is the same as the edition used for construction.

RAI 4.3.1-2: Table 4.3.1, URCS Design Transients" lists the number of transient cycles logged as of July 11, 2002, for the listed transients.

Indicate if the listed transients have been logged since the start of plant operation.

If logging of transients was implemented after the start of plant operation, state the basis for estimating the number of cycles prior to the initiation of logging.Response:

Table 4.3.1 lists the accrued design transients that have been incurred from the beginning of plant operation through July 11, 2002. The numbers of transient cycles occurring prior to the initiation of logging were determined through review of station logs.These cycles are included in the totals listed in Table 4.3.1.

Attachment 1 to 2CAN060403 Page 2 of 15 RAI 4.3.1-3: If additional transients have been logged since July 11, 2002, provide an updated Table 4.3.1 reflecting the most recently logged transients.

Response:

Major transient cycles that have occurred since July 11, 2002, are two reactor trips from low power associated with normal reactor shutdowns, one cooldown, and one heatup for a refueling outage. As indicated in Table 4.3.1, the numbers of projected transient cycles through the period of extended operation are well below the numbers of assumed design transient cycles. Consideration of transient cycles since July 11, 2002, does not change this result.RAI 4.3.1-4: In Table 4.3.1, the number of logged reactor trips from 100% transients are listed as 77.14, and the logged loss of reactor coolant flow at 100% transients are listed as 2.91. Provide an explanation why these are not whole numbers.Response:

These transients were assigned partial values if they did not occur from 100%power since transients initiated from lower power levels resulted in smaller temperature changes.RAI 4.3.2-1: Provide the ASME Section III Code edition that was used for non-Class 1 fatigue analyses.Response:

Non-Class 1 piping and in-line components were designed in accordance with ANSI B31.1 or ASME 1II, Classes 2 and 3 (Subsections NC and ND). Non-Class 1 pressure vessels, heat exchangers, storage tanks and pumps were designed in accordance with ASME Section VIII (Division I) or ASME Section 1II, Classes 2 and 3 (Subsections NC and ND). See Table 3.2-4 of the ANO-2 Safety Analysis Report (SAR) for a detailed listing of codes and standards for nuclear components at ANO-2.RAI 4.3.2-2: Provide the basis for the temperature screening criteria 220'F for carbon steel and 270 0 F for austenitic stainless steel in Section 4.3.2, 'Non-Class 1 Fatigue." Response:

The threshold value of 220'F for thermal fatigue of carbon steel piping is based on an initial ambient temperature of 70 0 F with a temperature increase of 150'F. The threshold value of 270 0 F for thermal fatigue of stainless steel piping is based on an initial ambient temperature of 70 0 F with a temperature increase of 200 0 F. The temperature differentials for carbon steel and stainless steel are based on industry sponsored (Electric Power Research Institute (EPRI) Report No. TR-1 04534, Fatigue Management Handbook, Volumes 1, 2 and 3, Research Project 3321, Revision 1, EPRI, December 1994)investigations and evaluations of thermal fatigue in operating nuclear power plant piping systems and components.

These screening criteria are consistent with the screening criteria used in the St. Lucie LRA, Section 4.3.2.RAI 4.3.2-3: The applicant states in Section 4.3.2 of the LRA: "Only the RCS hot leg sampling piping may exceed 7,000 cycles during the period of extended operation.

However, a calculation was revised to justify RCS sampling to occur at any reasonable frequency for 60 years of operation without exceeding the allowable number of cycles.Therefore, fatigue analyses for all non-Class 1 components at ANO-2 remain valid for the period of extended operation, in accordance with 1OCFR54.21(c)(1)(i)."

Attachment 1 to 2CAN060403 Page 3 of 15 1 OCFR54.21 (c)(1) states: .....The applicant shall demonstrate that -(i) The analyses remain valid for the period of extended operation;(ii) The analyses have been projected to the end of the period of extended operation From the description in the LRA paragraph above, if the analysis of the RCS hot leg sampling piping was revised and projected to the end of the period of extended operation, it should, therefore, fall in to the (ii) category.

To verify whether it does, and to resolve the apparent ambiguity in the LRA paragraph, the staff requests that the applicant clarify this statement by providing the following:

1. The value of the highest expansion stress range Se, the allowable stress range Sa for the piping material, and the number of expected thermal cycles, in the current licensing basis analysis.2. The projected number of cycles (less than 20,000) for the period of extended operation, and the basis for this number of cycles.3. The value of the highest expansion stress range Se and the allowable stress range Sa for the period of extended operation.

If Se was calculated from a revised thermal expansion analysis, provide the basis for this analysis.Response:

The RCS sampling line calculation was revised by assuming a lower stress allowable that corresponds to a stress range reduction factor of 0.8. The allowable number of cycles at the lower stress is 22,000. The number of thermal cycles projected for the sampling line through the period of extended operation is less than 20,000. Therefore, the RCS hot leg sampling line is qualified through the period of extended operation.

Additional details of the calculation are available for review on site.1. The RCS hot leg sampling line was initially analyzed to ASME Section I1I, Subsection NC rules. The basic design Code is USAS B31.7-1969 with Addenda through Summer 1971. The piping material for this line is Type 316 stainless steel resulting in an allowable stress range of Sa = 27,675 psi. The highest expansion stress range was determined to be Se = 24,910 psi. This analysis assumed less than 7,000 thermal cycles.2. At ANO-2, the RCS fluid is sampled four times per week and more often during startup and shutdown for outages. Conservatively assuming a sampling frequency of once per day and a 90% capacity factor, the number of cycles projected through the period of extended operation for the RCS sampling line is 365*60*0.9

= 19,710.3. The stress analysis for this line was revised to demonstrate acceptable operation.

The revision utilized a stress range reduction factor of 0.8. Per ASME Section I1I, NC-3652.3, the requirements of Equation (10) or Equation (11) must be met for acceptable thermal expansion stresses.

The revision utilized Equation (11), resulting in the RCS hot leg sampling line being qualified for up to 22000 cycles. Utilizing Equation 11 as detailed in ASME Section 1I1, NC-3650, the highest combined sustained plus thermal expansion load was determined to be Ste = 38,100 psi. This is within the allowable stress range of (Sh + Sa) = 38,840 psi. The highest expansion stress range, Se, was not changed for the revised analysis.

Attachment 1 to 2CAN060403 Page 4 of 15 RAI 4.3.3.3-1:

The NRC Safety Evaluation Report of CNP-387-P, included in CNP-387-P, Revision 1-P-A, recommends that licensees perform volumetric examination of the pressurizer elbow body and welds, as part of the ASME Section Xl inservice inspection program. State whether volumetric examination of the pressurizer elbow is currently included in the ANO, Unit 2, ASME Section XI inservice inspection program, and whether this will be continued during the period of extended operation.

Response:

The SER applicable to ANO-2 is Safety Evaluation for Combustion Engineering Owners Group Report CEN-387-P, Revision 1, Pressurizer Surge Line Thermal Stratification Evaluation (NRC Bulletin 88-11) (TAC No. M72109), dated July 23, 1993.Commitments regarding inspections at ANO-2 in response to NRC Bulletin 88-11 have been superseded by the ANO-2 risk-informed inservice inspection (RI-ISI) of ASME Class 1 piping, as approved by the NRC (References 1, 2, 3, and 4 below). Pressurizer surge line piping welds and elbow base metal selected for volumetric inspection based on risk category are identified in Reference 1, Table 10, Page 44, Elements16-003, 16-004,16-011, and 16-012. Volumetric examinations of pressurizer surge line piping welds and elbow base metal identified in Reference 1 will be carried forward through the period of extended operation.

Reference 2, NRC Question 3, Page 10, requested ANO-2 to address any changes in the current licensing basis and confirm that existing augmented examinations, such as IEB 88-11, inspections will not be impacted.

The ANO-2 response in Table 3-1 of Reference 2 is repeated below."The current ANO-2 ISI Program includes augmented examinations performed in response to various NRC issued bulletins and notice. The EPRI RI-ISI process defines an explicit set of attributes to consider in assessing the potential existence of a degradation mechanism.

The thermal fatigue and stress corrosion cracking concerns addressed in these NRC documents were inputs considered in the development of the EPRI degradation mechanism criteria.

As such, these concerns are inherently considered in the application of the EPRI RI-ISI process. Consequently, the RI-ISI program supercedes these augmented programs." The NRC acceptance of this statement is contained in the NRC SER of the ANO-2 RI-ISI submittal, Section 3.2.1. (Reference 4)

References:

1. Letter from Dwight C. Mims to USNRC, Arkansas Nuclear One -Unit 2, Docket 50-368, License No. NPF-6, Risk-Informed Inservice Inspection Pilot Plant Submittal forANO-2, dated September 30, 1997.2. Letter from Jimmy D. Vandergrift to USNRC, Arkansas Nuclear One -Unit 2, Docket 50-368, License No. NPF-6, Additional Information in Support of the Risk-Informed Inservice Inspection Pilot Application, dated October 8, 1998.3. Letter from Jimmy D. Vandergrift to USNRC, Arkansas Nuclear One -Unit 2, Docket 50-368, License No. NPF-6, Information to Support Risk-Informed Inservice Inspection Pilot Application, dated November 25, 1998.

Attachment 1 to 2CAN060403 Page 5 of 15 4. Letter from John N. Hannon (NRC) to C. Randy Hutchinson (EOI), Request to Use Risk Informed Alternative to the Requirements of ASME Code Section XI, Table IWX-2500 at Arkansas Nuclear One, Unit No. 2 (TAC No. M99756), dated December29, 1998.RAI 4.5-1: For the discussion of prestressing force losses over the initial 40 years, the LRA Section refers to SAR Section 3.8.1.3.4.

This section of the SAR discusses the design approach used in designing the containment to satisfy the load combinations in SAR Section 3.8.1.3.3.

There is no discussion of the estimation of projected prestressing forces after 40 years of operation.

As the estimated prestressing forces at 40 years and 60 years depend upon the regression analysis of these time dependent attributes (i.e., creep of concrete and relaxation of prestressing steel), please provide the estimated values of these attributes which were used in arriving at the minimum required prestressing forces.Response:

The estimated values of creep of concrete and relaxation of prestressing steel, used in the ANO-2 containment analysis calculations for 40 and 60 years are: Creep and shrinkage of concrete = 420 p in/in Relaxation of prestressing steel = 14.28 ksi for hoop, vertical and dome tendons RAI 4.5-2: The use of 1OCFR54.21(c)(1)(ii) and (iii) is appropriate for concrete containment tendon prestress time-limited aging analysis (TLAA). However, the staff needs to assess the plant specific operating experience regarding the residual prestressing forces in the containment.

Based on the analysis performed per lOCFR54.21(c)(1)(ii), the applicant is requested to provide the following information: (a) Minimum required prestressing forces for each group of tendons in terms of force per tendon.(b) Trend lines of the projected prestressing forces for each group of tendons based on the regression analysis of the measured prestressing forces (see NRC Information Notice 99-10 for additional information).(c) Plots showing comparisons of prestressing forces projected to 40 years and 60 years of operation, with the minimum required prestress for each group of tendons.The staff requests that the comparison curves be constructed in force per tendon as opposed to force per wire or strand since the acceptance criteria in Subsection IWL of Section Xl of the ASME Code uses these units. Furthermore, as stated in NRC Information Notice 99-10 the 'Calculation of the average effective wire forces in the tendon from the measured tendon force is made only to ensure that (the measured lift-off force) does not exceed 70% of the guaranteed ultimate tensile strength of the wire."

Attachment 1 to 2CAN060403 Page 6 of 15 Response: (a) The minimum required prestress wire forces for each group of tendons based on current site documentation is as follows (for 59 psig building design pressure).

The listed values in terms of forces per tendon were obtained by conservative method of multiplying the values from the attached curves by the total number of wires in each tendon which is 186.Hoop tendons 6.48 kips x186 = 1205.28 kips per tendon Dome tendons 6.63 kips x 186 = 1233.18 kips per tendon Vertical tendons 7.37 kips x 186 = 1370.82 kips per tendon (b) Trend lines of the projected prestressing forces for each group of tendons are attached.Prior to implementation of ASME Section Xl, Subsection IWE/IWL, ANO-2 took credit for results of the ANO-1 reactor building tendon inservice inspection, as allowed by code due to similarity of the two containments.

The curves are not based on a regression analysis per Information Notice 99-10. However, Entergy did evaluate its current method against the regression analysis outlined in Information Notice 99-10 during the ANO-1 15-year surveillance.

This analysis showed that the measured tendon prestress forces are well within the projected losses when compared against the original curve data. Entergy began using a random sampling software program for tendon selection in 1999 for the ANO-1 25-year and the ANO-2 20-year surveillances in accordance with the requirements of 10CFR50.55a.

Entergy used a design of 8%relaxation loss and a "normalized force" calculation to account for elastic losses during initial tensioning.

ANO has not experienced relaxation losses greater than expected during tendon surveillances.

The trending results for the three groups of tendons are provided below.

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., :_ ..-0 N3 >to 0 (a > a)CD z 0_L C)w 0) 3 0 0 (D_-, p_L 0 cn c') _L 0 Attachment 1 to 2CAN060403 Page 14 of 15 RAI 4.5-3: In Section A.2.2.4 of the SAR Supplement to the LRA, the applicant summarizes the results of this TLAA, and states, 'Calculation of the acceptability of the effective prestress of the containment building post-tensioning system at 60 years has been performed to show that the containment building tendon elements will be acceptable for the period of extended operation in accordance with 1OCFR54.21 (c)(1)(ii)." The Staff is requesting that the applicant enter the target values in the SAR Supplement.

Previous applicants have provided these values in (1) tabular form, (2) descriptive form, or (3) as an amendment to the plant's technical specifications with a reference to the technical specifications in their SAR Supplement.

Response: ANO-2's containment inservice inspection program (Section B.1.13), in accordance with ASME Section Xl, delineates the required documentation and the acceptance criteria for the prestress forces applicable for the period of extended operation.

The validity of the prestress analysis, per ASME Section XI, subsection IWL is demonstrated in site documents.

The adequacy of the aging management program (i.e., IWL) is assured since, as described in Section B.1.13 of the LRA, the program is consistent with the NUREG-1 801 program and with current regulatory requirements.

In accordance with the Statements of Consideration for the license renewal rule, the plant-specific licensing basis must be maintained during the renewal term in the same manner, and to the same extent, as during the original licensing term. Therefore, a summary table showing minimum required prestress forces for each group of tendons is not warranted in the SAR.RAI 4.6-1: Provide the loading conditions and corresponding transient cycles used in the fatigue analysis of the containment liner plate and penetrations.

Response:

The loading conditions for the containment liner plate and penetrations are provided in Section 3.8.1.3 of the SAR. Fatigue is not explicitly addressed in the containment analysis because the calculated peak stress intensities resulted in allowable fatigue cycles that far exceeded the projected number of anticipated cycles for all operating conditions.

If specifically addressed, the transient cycles would be similar to those identified in ANO-1 SAR Section 5.2.1.4.7.3.

RAI 4.6-2: Provide the ASME Section III cumulative usage factors and locations from the recently completed containment liner plate and containment penetration fatigue analyses, showing that these fatigue TLAAs will remain valid for the period of extended operation.

Response:

As stated in the previous response, the calculated peak stress intensities result in allowable fatigue cycles that far exceed the projected number of anticipated cycles for all operating conditions through the period of extended operation.

As a result, fatigue was not a controlling factor in the containment and liner plate analysis.

Therefore, the ASME Section III cumulative usage factors were not determined.

Due to the similarity between the two containments, if usage factors were specifically addressed, the transient cycles from the ANO-1 SAR Section 5.2.1.4.7.3 would be used as follows: Thermal cycling due to annual outdoor temperature variations

= 60 cycles for a 60-year life Thermal cycling due to reactor building interior temperature variations

= 500 cycles for 60-year life Thermal cycling due to design basis accident = 1 cycle Attachment 1 to 2CAN060403 Page 15 of 15 The containment was evaluated for the loading conditions shown in SAR Section 3.8.1.3.3.

The evaluation included an axisymmetric finite element analysis of the containment structure primarily targeting the areas of discontinuity at the ring girder and the haunch area. The finite element analysis identified a maximum liner plate strain value of -1381 E-6, for the maximum accident load condition, D+F+P+Ta.

Applying standard stress strain equations (i.e., stress = modulus of elasticity multiplied by strain or 30 E6 x -1381 E-6), maximum stress was determined to be approximately 42 ksi. Similarly, for normal conditions, D+F+To, for a maximum strain value of -979 E-6, the resulting stress is approximately 30 ksi.RAI 4.7.6-1: Provide a discussion indicating that the surge line fatigue TLAA was reevaluated to determine if additional pipe breaks need to be postulated at locations where the ASME Code Section III cumulative usage factor (CUF) may exceed the pipe break postulation criterion for Class 1 piping of 0.1 during the period of extended operation.

Response:

As described in Section 4.3.1 of the ANO-2 LRA, CUFs for the Class 1 components designed in accordance with ASME Section III were compiled and the RCS design transients used to develop the CUFs were identified.

The numbers of RCS design transients accrued through 2002 for ANO-2 were reviewed and these numbers were linearly extrapolated to 60 years of operation as reported in Table 4.3-1 of the ANO-2 LRA. In all instances the number of RCS transients assumed in the original design is greater than the number projected for 60 years of operation.

Therefore, the cumulative usage factors for pressurizer surge line items remain valid for the period of extended operation in accordance with 1OCFR54.21(c)(1)(i).

The RCS design transients are monitored through the fatigue monitoring program, which is discussed in Appendix B of the ANO-2 LRA.The limiting intermediate pressurizer surge line break locations are listed in Table 3.6-1 (A)of the ANO-2 SAR. The analytical method used to evaluate the dynamic response to postulated pipe breaks of the intermediate surge line locations is described in Section 3.9.1.7 of the ANO-2 SAR. Since the pressurizer surge line CUFs do not change, no new break locations are introduced.

Attachment 2 2CAN060403 List of Regulatory Commitments Attachment 2 to 2CAN060403 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE SCHEDULED ONE- eCc One) COMPLETION TIME CONTINUING DATE COMMITMENT ACTION COMPLIANCE (If Required)Volumetric examinations of pressurizer X July 17, 2018 surge line piping welds and elbow base metal identified in Reference 1 will be carried forward through the period of extended operation.

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