ML022560447

Response to Request for Additional Information Regarding Steam Generator Replacement & Power Uprate License Amendment Request

ML022560447
Person / Time
Site:	Palo Verde
Issue date:	09/06/2002
From:	Mauldin D Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-04837-CDM/TNW/RAB
Download: ML022560447 (29)

Text

David Mauldin 10 CFR 50.90 Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering TEL (623) 393-5553 P 0 Box 52034 Generating Station and §upport FAX (623) 393-6077 Phoenix, AZ 85072-2034 102-04837-CDM/TNW/RAB September 6, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-37 Washington, DC 20555

Reference:

1. Letter No. 102-04641-CDM/RAB, Dated December 21, 2001, from C. D.

Mauldin, APS, to U. S. Nuclear Regulatory Commission, "Request for a License Amendment to Support Replacement of Steam Generators and Uprated Power Operations"

2. Letter, Dated June 14, 2002 from J. N. Donohew, USNRC, to G. R.

Overbeck, "Palo Verde Nuclear Generating Station, Unit 2 - Request For Additional Information Regarding Power Uprate License Amendment Request (TAC No. MB3696)"

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 2, Docket No. STN 50-529 /

Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request In Reference 1, Arizona Public Service Company (APS) submitted a license amendment request to support steam generator replacement and uprated power operations for PVNGS Unit 2. In Reference 2, the NRC provided requests for additional information from the Mechanical and Civil Engineering Branch, the Reactor Systems Branch, the Materials and Chemical Engineering Branch, the Plant Systems Branch and the Probabilistic Safety Assessment Branch. to this letter provides written responses to the questions from the Mechanical and Civil Engineering Branch. Responses to questions from the remaining branches will be submitted separately.

Responses to questions 1, Ia, lb, 2, 4, 5 and 7 in Attachment 2 refer to Enclosure 1, which is considered proprietary information by the Westinghouse Electric Corporation.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance

• Wfr Callaway 0 Comanche Peak 9 Diablo Canyon 0 Palo Verde a South Texas Project 0 Wolf Creek

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request Page 2 Westinghouse requests that Enclosure 1 be withheld from public disclosure in accordance with 10CFR 2.790. Enclosure 2 contains an affidavit from the Westinghouse Electric Corporation stating the reasons that Enclosure 1 should be considered as a proprietary document.

No commitments are being made to the NRC in this letter.

Should you have any questions, please contact Thomas N. Weber at 623-393-5764.

Sincerely, CDM/TNW/RAB/kg Attachments:

1. Notarized Affidavit

2. Plant Systems Branch Questions and APS Responses

Enclosures:

1. Excerpts from the Impact of RSG and PUR on the Structural Integrity of the RCS Components and Supports

2. Affidavit from Westinghouse Electric Corporation Submitted Pursuant 10 CFR 2.790 to Consider Enclosure I as a Proprietary Document cc: E. W. Merschoff (NRC Region IV)

J. N. Donohew (NRC Project Manager)

D. G. Naujock (NRC Project Manager)

N. L. Salgado (PVNGS)

A. V. Godwin (ARRA)

Attachment 1 STATE OF ARIZONA )) SS.

COUNTY OF MARICOPA )

I, Paul F. Crawley, represent that I am Director, Nuclear Fuel Management, Arizona Public Service Company (APS), that the foregoing document has been signed by me on behalf of APS with full authority to do so, and that to the best of my knowledge and belief, the statements made therein are true and correct.

Paul F. Crawley" Sworn To Before Me This /-) Day Of ,C e, ,2002.

a Public nýýf

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b~IILE'AL" L Nora E. Meado No~ry Pubfio.Mo Notary Commission Stamp

)Westinghouse Proprietary Affidavit pursuant to 10 CFR 2.790 Page 1 of 2 I, Ian C. Rickard, depose and say that I am the Licensing Project Manager of Westinghouse Electric Company LLC (WEC), duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and described below.

I have personal knowledge of the criteria and procedures utilized by WEC in designating information as a trade secret, privileged, or as confidential commercial or financial information.

This affidavit is submitted in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations for withholding proprietary information and in conjunction with the application of Arizona Public Service Company for withholding this information. The information for which proprietary treatment is sought is contained in the following document that has been appropriately designated as proprietary:

Enclosure 1 to WEC letter LTR-OA-02-99, "Response to Questions 1, la, lb, 2, 4, 5, and 7 of the Request for Additional Information Regarding the PVNGS-2 Power Uprate (Proprietary and Non-Proprietary Versions)," July 29, 2002 Pursuant to 10 CFR 2.790(b)(4) of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information included in the document identified above should be withheld from public disclosure.

1. The information sought to be withheld from public disclosure is owned and has been held in confidence by WEC. It consists of data regarding component loads, stresses, stress intensities, cumulative usage factors, and related design information and analytical models provided in responses to NRC Requests for Additional Information (RAIs) concerning the Palo Verde Nuclear Generating Station Unit 2 (PVNGS-2) power uprate program.

2. The information consists of analyses or other similar data concerning a process, method or component, the application of which results in substantial competitive advantage to WEC.

3. The information is of a type customarily held in confidence by WEC and not customarily disclosed to the public..

4. The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.

5. The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements that provide for maintenance of the information in confidence.

6. Public disclosure of the information is likely to cause substantial harm to the competitive position of WEC because:

a. A similar product or service is provided by major competitors of WEC.

b. WEC has invested substantial funds and engineering resources in the development of this information. A competitor would incur similar expense in generating equivalent information.

Westinghouse Proprietary Affidavit pursuant to 10 CFR 2.790 Page 2 of 2

c. The information consists of data regarding component loads, stresses, stress intensities, cumulative usage factors, and related design information and analytical models provided in responses to NRC RAIs concerning the PVNGS-2 power uprate program. The availability of such information to competitors would enable them to design their product or service to better compete with WEC, take marketing or other actions to improve their product's position or impair the position of WEC's product, and avoid developing similar technical analysis in support of their processes, methods or apparatus.

d. Significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be included in pricing WEC's products and services. The ability of WEC's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.

e. Use of the information by competitors in the international marketplace would increase their ability to market comparable products or services by reducing the costs associated with their technology development. In addition, disclosure would have an adverse economic impact on WEC's potential for obtaining or maintaining foreign licenses.

Ian C. Rickard Licensing Project Manager Westinghouse Electric Company LLC Sworn to before me this 2 day of 2002 Notary comsinPubliciC/*

My commission expires:

Attachment 2 NRC Mechanical and Civil Engineering Branch Questions and APS Responses

Mechanical & Civil Engineering Branch NRC Question 1:

The Nuclear Steam Supply System (NSSS) at Palo Verde Nuclear Generating Station (PVNGS), Unit 2 was approved by the NRC staff via NUREG-0852, "Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System Combustion Engineering Standard Safety Analysis Report (CESSAR) System 80." The CESSAR describes the design of the reactor coolant system (RCS), its components, and their supports. The CESSAR describes the methodologies used to develop limiting loads and their locations, and also contains interface requirements between the CE-supplied System 80 NSSS and the rest of the plant. The PURLR, to the application implies that the analyses which support steam generator replacement (SGR) and power uprate (PUR) may be significantly changing the CESSAR methodologies and assumptions for Unit 2.

NRC Question 1 .a:

With respect to RCS stresses, including piping, components, supports, and tributary piping, provide a clear description of the methodologies used for determining the limiting stresses and cumulative usage factors (CUFs) for the SGR/PUR conditions. Describe any changes to the methodologies that were approved as part of CESSAR, and justify the acceptability of any methodology changes for showing compliance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (i.e., ASME Code) requirements. Describe any significant changes to the design transients that are used for the structural design of the NSSS. Also, discuss any changes to the interface requirements that resulted from the SGR/PUR.

APS Response:

The ASME Boiler and Pressure Vessel Code (B&PV) Code provided the basis for all RCS component stress evaluations. The original code year from the Analyses of Record (AORs) were maintained in all cases with the exception of the Replacement Steam Generators (RSGs). The Code of Record for Original Steam Generators (OSGs) is ASME, Sections II, Ill, V, and Xl, 1971 up to and including Winter 1973 Addenda.

The RSGs were designed and fabricated to the newer, up to date requirements of the 1989 Edition (no Addenda) of the ASME Code, Sections II, Ill, V, and Xl.

The engineering evaluations performed for the RSGs and operation at PUR conditions consisted of establishing new revised loads and comparing them with the ASME code allowables for each load case and component.

The only methodology changes from CESSAR are:

1) Implementation of reduced dynamic effects of non-main coolant piping breaks due to application of the NRC approved Leak-Before-Break (LBB) methodology to the Main Coolant Loop (MCL) piping, and

2) the use of the ANSYS computer code to replace the STRUDL (used for RCS and Component Seismic analyses), CEDAGS (used for RCS Loss of Coolant Page 1

Mechanical & Civil Engineering Branch Accident (LOCA)), and SAPIV (used for a new three-dimensional CEDM model) computer codes.

The new ANSYS computer models were benchmarked with the STRUDL and CEDAGS original models and found to be acceptable for the intended application.

Before performing the structural integrity analyses of the Reactor Vessel (RV),

Pressurizer, Reactor Coolant Pumps (RCPs), RCS piping and their supports, new Normal Operating Pressure (NOP), Seismic and LOCA (based on Branch Line Pipe Break (BLPB)/LBB methodology) loads on the RCS were determined for the RSG and operation at PUR conditions. New Dead Weight (DW) and thermal expansion loads (i.e., the load component comprising the NOP loads) were determined by applying conservative factors to the existing DW and thermal loads. Finite Element Analyses (FEA) using detailed RCS and structural building models were use to determine the new seismic and LOCA loads. Each model utilized appropriate boundary conditions. In the case of the LOCA analyses, the component and piping support gaps were also modeled, which required using non-linear FEAs. The seismic and LOCA analyses were based on the original licensing basis time history analysis methods, and the newly generated BLPB type forcing functions for simulated breaks involving the Main Steam Lines (MSLs), Feedwater (FW) lines, Shutdown Cooling (SDC) lines, Safety Injection (SI) lines, and the pressurizer surge line. These models utilized the ANSYS computer code.

All new component loads were tabulated and compared with the original component loads. In most cases, the original design loads remained bounding. For the few instances where new loads exceeded those of the AORs, new design loads (including nominal margins of 20%) were generated and incorporated into the component specifications. The new loads were either reconciled or re-analyzed, compared to the ASME code allowables to verify acceptability, and were documented in calculations and in the summary form in the Design Report Addenda for each affected component. For components where the original design loads did not change or remained bounding, both the component specification and Design Report were amended to indicate that the effects of RSG/PUR were examined and that there was no impact on the AORs.

The as-calculated loads and motions, for RSG and PUR operation, throughout the RCS formed the basis for determining conservative design loads for the various RCS components and their supports. It was concluded that normal operating conditions (pressure and temperature) did not change significantly from the original design used in the AOR. In addition, the effects of RSG/PUR on Anticipated Operational Occurrences (AOOs) were examined. It was determined that the original AORs for AOOs remained bounding for operation at PUR conditions. Therefore, the existing fatigue evaluations and resulting CUFs were not affected, and the existing AORs remained valid.

Asymmetric loadings on the RV head that result from the Control Element Drive Mechanisms (CEDMs) and RV head service structure were addressed using LBB methodology. This methodology was also used to address the In-Core Instrumentation (ICI) guide tubes and the ICI nozzles that are located at the bottom of the RV. BLPB type LOCA loads were applied to the MCL piping, RCPs, and their associated supports.

The faulted loads used in the AORs remain bounding.

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Mechanical & Civil Engineering Branch Interfacing loads with the containment building changed slightly. The RV supports were unaffected. The vertical loads on the Steam Generator (SG) sliding base increased due to a higher RSG weight. The effect of the higher loads on the stresses in the SG vertical support structures (pads, sliding base, bolting, key ways etc.) was verified to be acceptable. The horizontal loads from the RSG into the snubber/linkage system were determined both for Seismic Operating Basis Earthquake (OBE) and combined Seismic Safe Shutdown Earthquake (SSE)/LOCA conditions. The new snubber loads were below the original design loads. The new loads into the lateral key ways were increased somewhat over the originally calculated as-built loads, but were still bounded by the original design loads. , Excerpts from the Impact of RSG and PUR on the Structural Integrity of the RCS Components and Supports for Palo Verde Nuclear Generating Station, provides additional information regarding RCS components and supports. This enclosure is Westinghouse Electric Co. proprietary information and should be withheld from public disclosure pursuant to 10 CFR 2.790.

NRC Question 1.b:

The application indicates that leak before break (LBB) is being utilized for the design of more components than discussed in the CESSAR or in the current licensing basis for Unit 2. Describe and justify the new applications of LBB and/or changes in the postulated break locations. Discuss whether these applications of LBB are based on a generic staff safety evaluation, or whether they are changes to the licensing basis that require specific NRC staff review and approval. Also, provide additional information on the continued applicability of LBB for the SGR/PUR conditions (i.e., evaluate the SGR/PUR condition against the criteria evaluated in Supplement 3 to NUREG-0852, including the margin between the leakage-size crack and the critical-size crack, and the material properties of the replacement steam generators (RSGs), replacement cold leg elbows, and associated field welds).

APS Response:

In accordance with 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 4, that was in effect during original plant design, the mechanical design basis for the original RCS configuration included postulated breaks in all high energy piping greater than one inch (1") in diameter. The NRC granted a partial exemption from the requirements of 10 CFR Part 50, Appendix A, GDC 4 to APS to allow the application of LBB methodology to MCL piping to eliminate pipe whip restraints and jet shields that were previously needed to mitigate the effects of Main Coolant Loop Breaks (MCLBs). The exemption was granted in:

Letter from G. W. Knighton, NRC, to E. E. Van Brunt, APS, Docket No. 50 529/530, November 29, 1985, Request for a Schedular Exemption from a Portion of General Design Criterion 4 of Appendix A to 10 CFR 50 Regarding the Need to Analyze Large Primary Loop Pipe Ruptures as a Structural Design Basis for Palo Verde Nuclear Generating Station Units 2 and 3.

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Mechanical & Civil Engineering Branch The technical bases that demonstrated that guillotine type failures of the RCS main loop piping need not be considered in the design basis for CE designed plants was provided to the NRC in the following letters:

"* Letter from A. E. Scherer, CE, to Darrell G. Eisenhut, NRC, Docket No. STN 50-470, June 14, 1983, with enclosure, "Basis for Design of Plant Without Pipe Whip Restraints for RCS Main Loop Piping" and

"* Letter from A. E. Scherer, CE, to Darrell G. Eisenhut, NRC, Docket No. STN 50-470F, December 23, 1983, with enclosure, "Leak Before Break Evaluation of the Main Loop Piping of a CE Reactor Coolant System," Revision 1, November 1983.

The NRC staff has previously reviewed and approved these analyses in:

• Letter from Cecil 0. Thomas, NRC, to A. E. Scherer, CE, Docket No. STN 50 470, October 11, 1984, with enclosure, "Safety Evaluation Report on the Elimination of Large Primary Loop Ruptures as a Design Basis."

In addition, the NRC has review and approved the application of LBB methodology to the PVNGS units in Section 3.6.2 of NUREG-0857, Safety Evaluation Report Related to the operation of Palo Verde Nuclear Generating Stations, Units 1, 2, and 3, through Supplement 12.

APS applied LBB methodology on a limited basis to allow removal of selected component and piping snubbers. APS increased the scope of the application of LBB methodology during the engineering and design phase of RSG/PUR to include all RCS tributary and MCL nozzles. This increased scope is within the scope of the approved LBB methodology. In addition,,the MCL piping, including the RSGs and replacement cold leg elbows, was evaluated at PUR conditions and compared to the above references to ensure that application of LBB methodology remained valid for the increased scope of application.

For RSG/PUR, LBB methodology was applied to all primary side branch line piping that interfaces with the RCS. The bounding BLPBs with respect to the RCS response, are breaks in the largest tributary pipes of:

"* main steam line,

"* feedwater line,

"* pressurizer surge line,

"* safety injection line, and

"* shutdown cooling line.

The terminal end breaks in all five piping systems listed above plus intermediate breaks in the main steam and SI lines remained as controlling pipe breaks. The limiting BLPBs that are considered for RSG/PUR are listed in Table 1.b-1.

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Mechanical & Civil Engineering Branch Table 1.b-1: Pipe Break Description 1 FW terminal end (1A quadrant) 2 MSL terminal end (1A quadrant) 3 MSL intermediate (1A quadrant), X thrust 4 MSL intermediate (1A quadrant), Z thrust 5 Surge line (HL 1) terminal end 6 SDC (HL 1) terminal end 7 SI 1A terminal end 8 S 1 B terminal end 9 SI 1B intermediate break The following paragraphs provide information regarding RCS piping cracks and replacement materials:

Leakage Detection and Crack Stability The areas of concern are the RV inlet and outlet nozzles since these are regions of relatively high stress. The SG inlet and outlet nozzles are of lesser concern.

For the RSG/PUR configuration, the bending moments for the normal operating loads at both the RV outlet and inlet nozzles are smaller than the original bending moments used to determine the leakage crack length (see the Table 1 .b-2). Consequently, the leakage crack under normal operating mechanical loads for the RSG configuration and operation at PUR conditions is longer than the leakage crack for the OSG configuration.

The leakage crack is defined as a crack that will leak 10 gpm at normal operating primary system pressure.

Table 1.b-2: NOP Bending Moment Comparison (ft-kips)

Location Original Values Used for RSG and PUR Leakage Crack Determination Configuration RV Outlet Nozzle 6,406 6,049 RV Inlet Nozzle 2,175 790 A 10 gpm leak at the RV outlet nozzle requires a crack that is approximately 8.5% of the pipe circumference. At the RV inlet nozzle under the same conditions, a 10 gpm leak requires a crack that is approximately 11% of the pipe circumference. This means that with the addition of any mechanical load, the crack length required to open a detectable crack will be shorter than when subjected to system pressure alone. Application of LBB methodology demonstrated that a critical crack of 50% of the pipe circumference remains stable when subjected to both normal operating and SSE loads. This confirms that the leakage crack length for the RSG configuration is below the stability criterion used in the LBB methodology.

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Mechanical & Civil Engineering Branch The normal operating and SSE loads used in the determination of crack stability for the OSG configuration and the normal operating and SSE loads for the RSG configuration are listed in Table 1.b-3. The combined normal operating and SSE loads for the RSG configuration are less than those used for the OSG configuration. Therefore, a leaking crack in the RV outlet nozzle or inlet nozzle occurring after RSG/PUR will be detectable well before the crack can grow to an unstable length.

Table 1.b-3: (NOP + SSE) Bending Moment Comparisons (ft-kips)

Location OSG Configuration RSG and PUR Configuration RV Outlet Nozzle 6,853 6,535 RV Inlet Nozzle 1,831 934 Replacement Piping, RSGs, and Weld Material Properties The results of the evaluation of primary piping, RSGs, and weld material properties to be used for the LBB analysis are summarized as follows:

(1) The stress-strain curve used in the original LBB evaluation provides a reasonable representation of the nominal stress-strain properties for the MCL piping base, RSGs, and weld materials considered.

(2) The J-R Curve for SA-516 Grade 70 plate was a good lower bound estimate for plate material. However, some weld metals tend to have an even lower toughness property. The lower bound SA-516 weld metal curve is considered a more appropriate lower bound for the MCL piping, RSGs, and weld materials being considered in this evaluation.

(3) The original LBB analysis resulted in an acceptable margin when the measured toughness properties were degraded by a factor of four. Since the weld metal lower bound toughness properties are higher than one fourth of the toughness properties used in the original analysis, the original analysis remains conservative and valid for the lower bound weld metal J-R Curve.

In addition, the replacement RCS piping is at the RSG outlet nozzles, which are not at the critical stress locations used in the LBB analysis.

The short-term LOCA-related mass and energy releases are used as inputs to the subcompartment analyses that are performed to ensure that the walls of a subcompartment can maintain their structural integrity during the short pressure pulse (generally less than 3 seconds) that accompanies a high energy line rupture within that subcompartment. The evaluated subcompartments include the SG compartment, the reactor cavity region, and the pressurizer compartment. For the SG compartment and the reactor cavity region, LBB methodology was used to qualitatively demonstrate that any changes associated with operation at PUR conditions are offset by the LBB benefit of considering smaller RCS nozzle breaks. The current licensing bases for these compartments remains bounding.

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Mechanical & Civil Engineering Branch The pressurizer subcompartment analysis assumes a double-ended guillotine break of the pressurizer surge line. The energy release rates were calculated for operation at PUR conditions and compared with the original plant design conditions. The original plant design energy release rates continue to bound the PUR plant energy release rates by approximately 10%. Since the AOR determined that the pressurizer subcompartment was adequate, and the energy release rates for the PUR condition remain bounded by the original energy release rates, the pressurizer subcompartment remains structurally acceptable.

The calculated peak internal containment pressure, which does not credit LBB, increases to 58 psig as a result of operation at PUR conditions and RSGs, but remains bounded by the original containment internal design pressure of 60 psig.

NRC Question 2:

For the RCS piping (including pressurizer (PZR) surge line and tributary piping),

components (including reactor vessel (RV), reactor coolant pumps (RCPs), RSGs, and PZR), and supports, provide the calculated maximum stresses and CUFs at the critical locations. Include the ASME Code allowable limits and the ASME Code edition and addenda used in the evaluation of the SGR/PUR conditions. If different from the ASME Code of record, provide a justification.

APS Response:

The RCS piping (including pressurizer and RCS surge line nozzles and tributary line nozzles), components (including RV, RCPs, and pressurizer), and supports, were evaluated to address the effects of the RSGs and operation at PUR conditions. provides the calculated maximum stresses and CUFs for RCS piping, components, and supports at the critical locations.

The original ASME Code of Record for evaluation of the RCS tributary piping and pressurizer surge line is ASME Section III, 1974 Edition, including the Winter 1975 addenda (Summer 1979 addenda for Subsections NB-3650 through 3680) and is documented in UFSAR Table 3.9-5. The evaluation of RCS tributary piping and pressurizer surge line to support operation at PUR conditions with RSG were performed using the original ASME Code of Record.

Table 2-1 is a summary of the revised Class 2 stress analysis performed for MCL tributary piping and the pressurizer surge line. This table provides the maximum calculated stress for each ASME defined condition and the associated ASME allowable stress. Table 2-2 provides the Class 1 maximum and allowable stresses, and CUFs that were calculated in the original AOR. The original AOR bounds the stresses associated with RSG/PUR.

The RSGs have been designed and fabricated to the new, up to date requirements of the ASME Code 1989 Edition with no addenda. The OSG were designed and fabricated to the requirements of ASME Code, 1971 Edition including all addenda through Winter 1973. The differences between the code of design and fabrication and the Code of Record will be reconciled before installation of the RSGs as required by ASME sections NCA 1140 and NCA 3220.

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Mechanical & Civil Engineering Branch The stresses at the critical component locations and CUFs for the RSGs are summarized in Table 2-3.

Table 2-1: Maximum Structural Stresses in Tributary Piping and Pressurizer Surge Line, Class 2 Equation 8 Equation 9B Equation 9D Equation 10/11 Piping Calculated Allowable Calculated Allowable Calculated Allowable Calculated Allowable Segment (psi) (psi) (psi) (psi) (psi) (psi) (psi) (psi)

CVCS Charging Line 10,313 15,900 17,284 19,080 27,835 38,160 25,752 27,475 CVCS-hRCS Loop Drains 7,706 15,900 12,411 21,816 31,321 38,160 34,701 43,375 CVCS Auxiliary Spray 10,684 15,900 17,762 19,080 25,791 38,160 25,652 26,888 Lines Pressurizer 4,741 15,900 4,801 19,080 8,041 38,160 16,198 27,475 Surge Line_________

Pressurizer Spray Line 10,684 15,900 17,762 19,080 25,791 38,160 25,652 26,888 SDC Lines 8,775 15,900 18,171 19,080 31,464 38,160 40,877 43,375 SI Loops 1A 10,327 18,600 15,051 22,320 22,540 39,360 27,082 27,600 and 1B SI Loops 2A 10,864 18,600 19,394 19,680 33,116 39,360 27,935 28,150 and 2B I I I I I Table 2-2: Class 1 Maximum Stresses and CUFs from original AOR Tributary Piping and Pressurizer Surge Line Equation 9 Equation Equation Equation Allowable CUE System Calculated Allowable 10 12 13 3.0 x Smr CVCS Aux Spray Line 15,468 24,225 56,147 8,793 44,629 48,450 0.5137 Maximum stress CVCS Charging Line Maiu Starg ies 17,362 27,000 67,038 8,178 38,630 54,000 0.3204 Maximum Stress CVCS C nLetow Line 212,673 24,225 62,332 40,384 22,243 48,450 0.2302 Maximum Stress CV C S - Letd ow n Lin e2 23 0 2 , 35 6 9 02 Maximum Stress2230 2,3 5690 2,0 , 04 4300, 00 84 0 8400.9 0. 9 Pressurizer Spray Line 20,716 24,225 61,817 26,150 47,255 48,450 0.6029 Maximum Stress SDC - Loop 1 Maximum 23,792 29,025 73,095 42,415 53,587 58,050 0.257 Stress I________________ I_____ I____ I_______

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Mechanical & Civil Engineering Branch Table 2-2: Class 1 Maximum Stresses and CUFs from original AOR Tributary Piping and Pressurizer Surge Line Equation 9 Equation Equation Equation Allowable CUE System Calculated Allowable 10 12 13 3.0 x Sm SDC - Loop 2 Maximum 13,843 29,025 68,304 49,413 44,646 56,100 0.1573 Stress SI Maximum Stress Loop 17,130 29,030 68,290 19,630 39,360 58,050 0.1819 1A SI Maximum Stress Loop 13,840 24,900 49,310 0 0 58,050 0.1854 1B SI Maximum Stress Loop 16,190 25,140 98,095 28,454 57,848 58,050 0.753 2A SI Maximum Stress Loop 19,953 29,025 86,446 28,454 57,777 58,050 0.274 2B Pressurizer Surge Line 13,677 24,000 58,319 41,700 39,461 48,564 0.495 Maximum Stress 13,677 24,000 I3 41,0 I Pressurizer Surge Line CUF Including Stratification 0.937 Table 2-3: Maximum Stresses and CUFs for RSG Components Maximum Allowable Stress (psi) Stress (psi) Stress Ratio Component Stress Category (A) (B) (A)/(B) CUF Pm 26,500 Sm= 26700 0.99 Tubesheet and Primary Head P, + Pb 38,300 1.5 Sm = 40,050 0.96 0.996 Pi + Pb + Q 79,800 3.0 Sm= 80,100 0.997 Pm 24,300 Sm = 26,700 0.91 Primary Inlet Nozzle P1 + Pb 29,900 1.5 Sm = 40,050 0.75 0.0416 P1 + Pb + Q 47,600 3.0 Sm = 80,100 0.59 Pm 17,000 Sm = 26,700 0.64 Primary Outlet Nozzle PI + Pb 23,600 1.5 Sm = 40,050 0.59 0.017 Pi + Pb + Q 42,400 3.0 Sm = 80,100 0.53 Pm 5,900 Sm = 26,700 0.22 Primary Manway Pi + Pb 22,900 1.5 Sm = 40,050 0.57 0.037 P1 + Pb+ Q 41,100 3.0 Sm = 80,100 0.51 Stud Replace every 6 (Six) years.

Pm 8,600 Sm = 26,700 0.37 Primary Divider Plate P1 + Pb 30,500 1.5 Sm = 40,050 0.87 0.08 PI + Pb + Q 64,500 3.0 Sm = 80,100 0.92 Pm 19,100 Sm = 26,700 0.72 Support Skirt PI + Pb 23,600 1.5 Sm = 40,050 0.59 0.155 PI + Pb + Q 79,400 3.0 Sm = 80,100 0.99 Page 9

Mechanical & Civil Engineering Branch Table 2-3: Maximum Stresses and CUFs for RSG Components Maximum Allowable Stress (psi) Stress (psi) Stress Ratio Component Stress Category (A) (B) (A)/(B) CUF Tube to Tubesheet Weld Pm 22,400 Sm = 26,700 0.96 0.668 Pm 20,400 Sm = 26,600 0.77 Tubes and Tube Upper Supports PI + Pb 20,400 1.5 Sm = 39,900 0.51 0 P1 + Pb + Q 36,200 3.0 Sm = 79,800 0.45 Pm 25,350 Sm = 26,700 0.949 Secondary Shell P1 + Pb 31,600 1.5 Sm = 40,050 0.789 0.009 PI + Pb + Q 39,000 3.0 Sm = 80,100 0.487 Pm 14,600 Sm = 18,250 0.8 Economizer FW Nozzle PI + Pb 24,000 1.35 Sm - 24,640 0.98 0.981 Pi + Pb + Q 79,800 3.0 Sm= 80,100 0.996 Pm 13,700 Sm = 21,500 0.64 Downcomer Blowdown Nozzle PI + Pb 30,400 1.5 Sm = 32,300 0.94 0.255 Pi + Pb + Q 54,400 3.0 Sm = 64,500 0.84 Pm 14,000 Sm = 23,330 0.6 Downcomer FW Nozzle P1 + Pb 19,300 1.5 Sm = 24,600 0.78 0.996 P1 + Pb + Q 58,700 3 Sm = 63,900 0.92 Pm 11,800 Sm = 18,200 0.65 Recirculation Nozzle P1 + Pb 28,400 1.5 Sm = 40,000 0.71 0.107 PI + Pb+ 47,500 3 Sm 80,100 0.59 Pm 22,500 Sm = 26,700 0.84 Steam Outlet Nozzle P1 + Pb 27,000 1.5 Sm = 40,000 0.67 0.169 P1 + Pb + Q 38,400 3 Sm = 80,100 0.48 Secondary Shell Instruments Pm 2,000 Sm = 18,200 0.11 Exempt SecondaryShellInstrument PM+ Pb+ Q 22,900 3 Sm = 55,100 0.41 Primary Head Instruments Pm 9,000 Sm = 23,300 0.39 Exempt PrimaryHeadInstruments Pm + Pb + Q 63,700 3 Sm = 69,900 0.91 Tubesheet Blowdown P,+ Pb 26,000 1.5 Sm = 31,600 0.82 Exempt TubesheetBlowdownPm + Pb + Q 69,900 3 Sm = 69,900 1 Secondary Manway P1+ Pb 32,600 1.5 Sm = 40,000 0.81 0.128 SecondaryManwayPi + Pb + Q 58,600 3 Sm = 80,100 0.73 Bolt 0.771 PM 26,300 Sm = 26,700 0.97 Secondary Handholes P, + Pb 32,500 1.5 Sm = 40,000 0.81 0.944 PI+Pb+Q 68,900 3 Sm = 80,100 0.86 Stud Replace studs every eighteen years Pm 12,300 Sm= 23,300 0.46 Upper Support Lugs P1 + Pb 27,900 1.5 Sm = 40,000 0.7 PI + Pb + Q 46,800 3 Sm = 80,100 0.58 Page 10

Mechanical & Civil Engineering Branch Table 2-3: Maximum Stresses and CUFs for RSG Components Maximum Allowable Stress (psi) Stress (psi) Stress Ratio Component Stress Category (A) (B) (A)/(B) CUF Dryers Assembly Design P, + Pb 3,500 28,700 0.12 Exempt Shear 5.9 20,000 0.3 Dryers Assembly Design PI+ Pb 25,800 73,500 0.35 Exempt Level D Shear 28,800 42,000 0.69 Separators Design Pm 550 12,600 0.04 Exempt P1 + Pb 4,100 28,700 0.14 Separators Design Pm 740 40,600 0.12 Exempt Level D PI + Pb 42,900 73,500 0.58 Pm 7,200 19,100 0.38 Exempt Shroud Assembly Design PM+ Pb 19,200 28,700 0.67 Shroud Assembly Design Pm 8,700 49,000 0.18 Exempt Level D Pm + Pb 29,100 73,500 0.4 Eggcrate Assembly Design Pm 8,700 18,100 0.48 Exempt P1 + Pb 0 21,100 0 Eggcrate Assembly Design Pm 29,600 40,600 0.73 Exempt Level D P1 + Pb 24,400 58,500 0.42 Downcomer FW Piping Assembly Design Pm+ Pb 8,700 27,100 0.32 Exempt Downcomer A Pm+ Pb 19,000 47,500 0.4 0.125 Level N/B ______

Downcomer FW Pm + Pb 1,700 63,000 0 Exempt Level Diruno+P+ 50 57,500 0.91_ 0.961 FW Distribution Box P + Pb +Q i52,500 57,500 0.91 0.961 NRC Question 3:

Section 5.3.3.1 of the PURLR indicates that the response spectrum for the containment basemat in the vertical direction for the operating basis earthquake is not bounded by the analysis of record. Provide an evaluation of the containment basemat stresses for this condition.

APS Response:

Section 5.3.3 of the PURLR describes the RV ICI Tubes. The wording of Section 5.3.3.1 should more appropriately read:

The ICI tubes are attached to the RV bottom head and terminate at the ICI seal table.

The original analysis of the ICI tubes was based on a set of "Configuration Spectra" that was intended to envelop both the RV configuration spectra and the containment basemat spectra. RSG/PUR did not affect the containment basemat spectra. However, the RV configuration spectra were affected somewhat, which required reconciliation of Page 11

Mechanical & Civil Engineering Branch the new RV spectra. Reconciliation demonstrated that the new RV spectra is comparable to the original design spectra.

Review of the original design OBE spectra verified that the existing ICI configuration spectra completely envelop both the PUR configuration RV spectra and the containment basemat spectra for the X direction (horizontal and parallel to the hot leg) and the Z direction (horizontal and perpendicular to the hot leg). However, for the vertical (Y) direction, both the PUR configuration RV spectra and the original containment basemat spectra exceed the existing ICI configuration spectra at certain frequencies. Since total enveloping of the containment basement and RV vertical spectra could not be demonstrated, it was necessary to recalculate the response of the ICI tubes to PUR RV conditions and to the original basement spectra, and to verify that this response is bounded by the original/existing ICI configuration spectra.

The AOR provided the response for selected locations for each mode and each excitation direction. The response to the PUR configuration spectra was calculated by multiplying each AOR modal response by the ratio of the new to the original modal accelerations. The response for each excitation direction was calculated-by combining the modal responses in accordance with Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis. The total response of the ICI tubes was then calculated using the Square-Root-of-the-Sum-of-the-Squares (SRSS) of the response due to the X, Y, and Z directions. The load results demonstrated that the response of the ICI tubes to the OBE excitation for the PUR configuration is less than the response calculated for the existing configuration.

Therefore, all ICI OBE results for the existing ICI spectrum configuration are valid and remain bounding for operation at PUR conditions.

The original AOR containment basemat spectra and the motions of containment were not changed as a result of the RSG/PUR.

NRC Question 4:

For the RV internals provide the maximum calculated stresses and CUFs for the SGR/PUR condition. Include the ASME Code allowable limits used in the evaluation, and the ASME Code edition and addenda. If different from the ASME Code of Record, provide a justification.

APS Response:

The ASME Code from the AORs was maintained in all cases with exception of the RSG.

The Code of Record for OSGs is ASME Code,Section II, Ill, V, and Xl, 1971, up to and including Winter 1973 Addenda. The RSGs were designed and fabricated to the requirements of the new, updated 1989 Edition of the ASME Code, Sections II, Ill, V, and Xl, no Addenda.

Tables 4-1 and 4-2 provide the maximum stress versus allowable stress values for normal operation plus upset condition, and faulted design condition, respectively. Table 4-1 also includes the CUFs for the RSG/PUR condition. Supplemental information may be obtained from Enclosure 1.

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Mechanical & Civil Engineering Branch Table 4-1: RVI Stress Summary for RSG and PUR - Normal Operation plus Upset Design Condition Major Component Stress Maximum Allowable CUF (1.0 Assembly Category (5) Stress (psi) (psi) (') Allowable)

Core Support PM 5,082 16,100 Barrel (CSB) Upper Flange Pm + Pb 14,151 24,150 < 0.462 Components Pm + Pb + Q 29,099 48,300 Pm 5,278 16,100 Cylinders Pm + Pb 9,214 24,150 < 0.462 Pm + Pb + Q 21,694 48,300 PM 2,815 16,100 Lower Flange Pm + Pb 10,166 24,150 < 0.462 Pm + Pb + Q 23,541 48,300 Pm 7,717 14,490 Snubber to Cylinder Weld Pm + Pb 14,275 21,735 < 0.462 Pm + Pb + Q 25,785 43,470 CSB to Lower Support PM 2,814 14,490 Structure (LSS) Flexure Pm + Pb 3,947 21,735 < 0.462 Weld Pm+ Pb + Q 18,107 43,470 LSS Insert Pin Pm 2,520 43,300 < 0.078 Components PM + Pb 11,377 64,950 Main Beam to Short Beam Pm 953 14,490 M eam Pm + Pb 8,261 21,735 < 0.078 Weld Pm + Pb + Q 37,246 43,470 Pm 10,668 14,490 Main Support Beam Pm + Pb 14,418 21,735 < 0.078 Pm + Pb + Q 28,317 43,470 PM 6,104 14,490 Cylinder Pm + Pb 8,633 21,735 < 0.078 Pm + Pb + Q 27,788 43,470 Pm 964 14,490 Raised Bottom Plate Pm + Pb 12,036 21,735 < 0.078 Pm+Pb+Q 13,692 43,470 Page 13

Mechanical & Civil Engineering Branch Table 4-1: RVI Stress Summary for RSG and PUR - Normal Operation plus Upset Design Condition Major Component Stress Maximum Allowable CUF (1.0 Assembly Category (5) Stress (psi) (psi) (1) Allowable)

Upper Guide PM 10,534 16,100 Structure Upper Flange Pm + Pb 22,490 24,150 < 0.31 (UGS) Pm + Pb + Q 31,994 48,300 Components Pm 1,651 16,100 Lower Flange Pm + Pb 12,918 24,150 < 0.31 Pm + Pb + Q 22,833 48,300 Pm 1,413 16,100 Control Element Assembly 0 (CEA) Guide Tube24,150 Pm + Pb + Q 39,458 48,300 Guide Tube to Upper Pm 1,391 12,075 Guide Structure Support Pm + Pb 2,525 18,113 0 Plate (UGSSP) Weld Pm + Pb + Q 17,226 36,225 Pm 645 16,100 UGS Support Plate Pm + Pb 13,596 24,150 0.118 Pm + Pb + Q 37,949 48,300 Pm 705 16,100 Fuel Alignment Plate Pm + Pb 13,891 24,150 0.721 Pm + Pb + Q 24,270 48,300 Pm 421 12,075 0 Pm+ Pb 4,166 18,133 Tube to Fuel Alignment Pm++Pb Pb+ Q 4,166 36,2 Plate (FAP) Weld Pm 4,166 36,225 Internal Pm 9,825 16,100 Core Shroud Pm + Pb 22,838 24,150 N/A (3)

Structures Pm + Pb + Q 39,920 43,470 CEA Shroud Tubes Pm+ Pb 22,464 24,150 0.32 Pm + Pb + Q 38,073 48,300 CEA Shroud Tube to Web Pm + Pb 7,890 8,452 0.093 Weld Pm + Pb + Q 23,499 48,300 Page 14

Mechanical & Civil Engineering Branch Table 4-2: RVI Stress Summary for RSG and PUR - Faulted Design Condition Major Stress Maximum Allowable (psi)

Assembly Component Category (5) Stress (psi) (2)

CSB Pm 16,857 38,640 Components Upper Flange Pm + Pb 42,802 57,960 Pm 30,998 38,640 Cylinders Pm + Pb 42,814 57,960 PM 14,826 38,640 Lower Flange Pm + Pb 34,622 57,960 PM 10,314 34,776 Snubbers @ Shell Pm + Pb 12,260 52,164 PM 14,826 34,776 CSB to LSS Flexure Weld Pm + Pb 31,160 52,164 LSS PM 6,152 91,000 Components Insert Pin Pm + Pb 27,779 136,500 Main Beam to Short Beam Pm 24,633 34,776 Weld Pm + Pb 49,057 52,164 Pmn P 21,656 1663,7 34,776 Main Support Beam Pm, + Pb 24,657 52,164 Pm 33,890 34,776 Cylinder PM + Pb 41,432 52,164 Pm 2,673 34,776 Raised Bottom Plate Pm + Pb 34,113 52,164 UGS Pm 35,415 38,640 Components Upper Flange Pm + Pb 49,758 57,960 Pm 11,203 38,640 Lower Flange Pm + Pb 54,372 57,960 Pm 4,657 38,640 CEA Guide Tube Pm + Pb 12,125 57,960 Pm 5,974 28,980 Guide Tube to UGSSP Weld Pm + Pb 6,173 43,470 Pmn 2,199 38,640 UGS Support PlatePM2193,4 Pm + Pb 44,894 57,960 UGS Pm 2,418 38,640 Components Fuel Alignment Plate Pm + Pb 44,891 57,960 Pm 451 28,980 Tube to FAP Weld + Pb 5,076 43,470 Tube_toFAPWeldPm Page 15

Mechanical & Civil Engineering Branch Table 4-2: RVI Stress Summary for RSG and PUR - Faulted Design Condition Major Stress Maximum Allowable (psi)

Assembly Component Category (5) Stress (psi) (2)

Internal Pm 31,677 38,640 Structures Core Shroud Pm + Pb 57,747 57,960 N/A (4) .109 inch (4) .628 inch (4) 1CEA Shroud Assembly Notes for Tables 4-1 and 4-2:

(1) Allowable stress criteria defined in ASME B&PV Code,Section III, Division 1, Subsection NG, 1974 Edition without addenda.

(2) Allowable stress criteria defined in ASME B&PV Code,Section III, Division 1, Appendix F, 1974 Edition without addenda.

(3) For RSG/PUR, it was determined that the AOR bound the structural evaluation of the Core Shroud. The AOR did not calculate fatigue usage.

(4) The CEA Shroud is deflection-limited, rather than stress-limited, in the faulted condition.

(5) Stress categories are as defined below:

Pm = Primary membrane stress Pm + Pb = Primary membrane plus bending stress Pm + Pb + Q = Primary membrane plus bending plus secondary stress NRC Question 5:

For the control element drive mechanisms (CEDMs), the PURLR describes changes in the methodology for determining stresses and CUFs. Describe the benchmarking of the new methodology, and discuss the new methodology's acceptability for determining stresses and CUFs for the SGR/PUR condition. Provide the maximum calculated stresses and CUFs at the critical locations of the CEDMs for the SGR/PUR condition.

Include the ASME Code allowable stresses and the ASME Code edition and addenda used in the evaluation of SRG/PUR. If different from the ASME Code of record, provide a justification.

APS Response:

The changes in the methodology are the use of a new three-dimensional CEDM model, use of the ANSYS code, rather than SAPIV code (for the response spectrum analyses),

and inclusion of BLPB loads in the faulted loads.

The new CEDM model uses a sufficient number of nodes to accurately represent the dynamic characteristics of the CEDM nozzle components and to provide a detailed load response distribution throughout the CEDM structure. The model was benchmarked by comparison of the calculated natural frequencies and mode shapes to test data. The first, second, and third mode natural frequencies of the CEDM model with the longest nozzle length were calculated to be 2.33 Hz, 9.61 Hz, and 10.32 Hz. These calculated values compare favorably with experimentally measured values of 2.32 Hz, 9.2 Hz, and 11.6 Hz.

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Mechanical & Civil Engineering Branch The CEDM model was used for DW, seismic, and BLPB analyses for the CEDM nozzles. The calculated response loads were used to evaluate the CEDM components operability for the upset condition (NOP plus OBE results) and the faulted condition (NOP plus SRSS combination of SSE and BLPB) results. CEDM operability was verified by comparing the predicted maximum CEDM deflections due to the combination of SSE and BLPB with those shown to be acceptable by static testing. A maximum calculated combined deflection was used in this comparison for conservatism.

The CUFs for all CEDM components calculated in the AOR remain valid for RSG/PUR.

This is also true for the nozzle because the small change in the OBE stresses does not affect the fatigue evaluation. provides a more detailed description of the CEDM analysis, as well as tables that list the results of the analysis, design loads used in the AOR, and a figure showing the CEDM model.

The CEDM analysis results can be summarized as follows:

1) The NOP mechanical loads are comparable to the AOR.

2) The calculated OBE loads are similar to the AOR loads.

3) The faulted loads are the sum of the NOP loads with the SRSS of the SSE and BLPB loads. The calculated faulted loads are lower than those in the AOR because the SSE horizontal response spectra are less severe.

4) The calculated loads for all CEDM components are less than the design loads used in the AOR except for the NOP plus OBE load on the nozzle and the axial load on the upper pressure housing.

5) The most limiting load for all CEDM components is the NOP plus OBE load on the nozzle. The calculated bending moment increased from the design value of 111.7 in-kips to 112.1 in-kips. However, the calculated nozzle stress remained less than the ASME Code allowable.

6) The increase in the axial load on the upper pressure housing is not significant for two reasons. The axial load causes a minimal increase in stress, and the AOR included a conservative DW load that covers the difference between the calculated and design loads. Therefore, the existing AOR remains bounding for operation at PUR conditions.

As discussed above, the design loads were all greater than the calculated loads, except for the NOP plus OBE load on the nozzle. Therefore, except for the Level A and B stress in the nozzle, all stresses calculated in the AOR remain valid. The calculated Levels A and B stress intensity increased by 2.8% from the AOR. The calculated Level A and B general primary membrane stress intensity in the CEDM nozzle is 22.86 ksi, compared to an allowable of 23.3 ksi. The allowable stress is from the ASME Code, 1971 Edition with Addenda through the Winter of 1973, which is the ASME Code of Record.

NRC Question 6:

Discuss the potential for flow-induced vibration of the steam generator (SG) tubes due to various mechanisms, including, the fluid-elastic instability, in the RSG at the PUR Page 17

Mechanical & Civil Engineering Branch condition. Describe the analysis methodology, damping value of the tubes, and the computer code used in the analysis. Also, provide the results of the predicted vibration levels during the normal operating condition and the worst case transient condition, including the calculated fluid-elastic instability ratios. Explain whether the above analysis results are applicable to the degraded SG condition and why.

APS Response:

The potential for Flow-Induced Vibration (FIV) is minimized due to SG design. The FIV analyses were performed on selected tubes based on tube span parameters, such as frequency and mode shape of vibration, and fluid flow parameters, including flow velocity and fluid density. The selected tubes were modeled with the ANSYS finite element code in order to determine their natural frequencies and mode shapes. The finite element models represent the entire length of the tube above the tubesheet, and include the fluid entrance regions in the vertical tube spans and the fluid exit regions in the tube bends. The tube supports are represented by displacement boundary conditions, which model the following types of tube supports at the appropriate locations along the length of the tube:

"* the diagonal and vertical strip supports restrain the out-of-plane displacement of the tube,

"* eggcrate supports restrain the in-plane and out-of-plane displacement normal to the tube axis, and

"* vertical strip supports with horizontal bars above and below the tube restrain all the in-plane and out-of-plane displacements.

The FIV analysis considers both fluid elastic instability and random turbulent excitation mechanisms. The methodology employed in the evaluation of fluid elastic instability is based on the experimental results and analytical procedures described in:

"* Heilker, W.J., Vincent, R.Q., "Vibration in Nuclear Heat Exchangers Due to Liquid and Two-Phase Flow," Engineering for Power, April 1981, Vol. 103, No. 2.

"* Heilker, W.J., Beard, N.L., Park, J.Y., "Flow Induced Vibration Analysis in Support of Design of the Yonggwang Units 3&4 Steam Generators," Proceedings of the International Symposium of Pressure Vessel Technology and Nuclear Codes and Standards, April 19-21, 1989.

"* Conners, H.J., Jr., "Fluidelastic Vibration of Heat Exchanger Tube Arrays," ASME Publication 77-DET-90, 1977.

• Connors, H.J., Jr., "Fluidelastic Vibration of Tube Arrays Excited by Nonuniform Cross Flow", Flow-Induced Vibration of Power Plant Components, ASME, PVP 41,1980, p.93.

The random turbulent excitation was evaluated using the methodology outlined in Article N-1340, Appendix N of the 1995 Edition of the ASME B&PV Code Section III and in Welding Research Council Bulletin No. 372. The damping values used in both of these evaluations vary with mode frequency and were obtained from tests of prototypically supported tubes. All FIV calculations include thermal-hydraulic results from the ATHOS code, using the appropriate tube support arrangement.

Page 18

Mechanical & Civil Engineering Branch The evaluation results showed a maximum Stability Ratio of 0.38 in the fluid exit bend region of the bundle. This result is below the design goal of 0.75. The turbulent displacements are also within the limit of 10 mils Root Mean Squared (RMS). The results reflect the beneficial effects of the design changes in the tube support system of the RSG. Results of the predicted vibration levels are not provided since the evaluation results demonstrated favorable stability ratios for normal and worst case transient conditions.

NRC Question 7:

Describe any changes to the thermal stratification of the PZR surge line and any changes to the thermal fatigue of the pressurizer spray nozzle.

APS Response:

The impact of operation at PUR conditions on the thermal load and transients for the pressurizer surge line and the spray line/nozzle were reviewed. The AORs remain bounding.

NRC Question 8:

Describe the methodology used to evaluate the balance of plant (BOP) piping, components (including pumps, valves, and heat exchangers), and supports. Justify differences from the original design methodology. Also, provide the calculated maximum stresses for the critical BOP piping systems. Include the ASME Code edition and addenda and ASME Code allowable limits. If different from the ASME Code of record, provide a justification.

APS Response:

The Balance of Plant (BOP) piping, components, and supports were originally evaluated using ASME Code, 1974 Edition, including the Winter 1975 addenda, as documented in UFSAR Table 3.9-5. The affected BOP piping, components, and supports were evaluated using the same methodology. Stress limits, as stated in Table 3.9-6 of the UFSAR, were used in BOP piping and component evaluations and stress limits, as stated in UFSAR Table 3.9-11, were used in BOP support evaluations.

There were no changes to the original ASME Code of Record for the BOP piping. The calculated maximum stresses in the affected BOP piping are provided in Table 8-1 for each ASME defined condition evaluated (i.e., normal, upset, emergency, and faulted).

The BOP heat exchangers, pumps, and valves that are most affected by RSG/PUR are those subcomponents of the main steam system. These include main steam safety valves (MSSVs), Main Steam Isolation Valves (MSIVs), turbine stop valves, turbine throttle valves, low pressure FW heaters, high pressure FW heaters, condensate pumps, FW pumps, and heater drain pumps. These components were evaluated at the predicted PUR mass flow rates, temperatures, and pressures. The original component design requirements were found to bound operation at PUR conditions.

Page 19

Mechanical & Civil Engineering Branch Table 8-1: Maximum Stresses In BOP Piping Systems Equation 8 Equation 9B Equation 9D Equation 10/11 Calculated Allowable Calculated Allowable Calculated Allowable Calculated Allowable Piping Segment (psi) (psi) (psi) (psi) (psi) (psi) (psi) (psi)

SG Blowdown 9,694 15,000 17,298 18,000 33,366 36,000 22,037 22,500 Recirculation FW &

Downcomer 9,778 15,000 16,130 18,000 22,338 36,000 9,511 22,500 Main Steam 5,991 17,500 12,612 21,000 26,624 42,000 9,053 26,250 NRC Question 9:

The PUR results in an increase in the main steam flow and the feedwater flow. Discuss the potential for flow-induced vibration in the main steam and feedwater piping and the BOP heaters and heat exchangers following PUR. Also, clarify whether vibration monitoring, consistent with OM-3, will be included in the startup testing program for PUR.

APS Response:

All secondary side component steam and water operating velocities are predicted to remain below the original component design limits when operating at PUR conditions.

Post-PUR startup testing will be conducted in accordance with the requirements of ASME OM-S/G-2000, "Standards and Guides for Operation and Maintenance of Nuclear Power Plants", including Part 3, "Requirements for Preoperational and Initial Start-up Vibration Testing of Nuclear Power Plant Piping Systems," which supercedes ANSI/ASME/OM-3.

NRC Question 10:

The SGR/PUR increases the post-accident containment temperature and pressure.

Discuss the effects of the SGR/PUR on the overpressurization of isolated piping segments (reference: Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions").

APS Response:

The calculated post-LOCA peak containment temperature and the time temperature profile have increased slightly. The slight increase in temperature has a minimal effect on equipment in containment, and the existing analyses performed to address Generic Letter 96-06 remain bounding for operation at PUR conditions. The systems and components within containment subject to post accident environmental heating and pressurization remain capable of withstanding the predicted peak pressures and temperatures.

Page 20

Mechanical & Civil Engineering Branch NRC Question 11:

Confirm whether the SGR/PUR will increase the accident sub-compartment temperature and pressure that affect the design basis for steel and concrete in the containment. If the structural steel and concrete will be affected, provide the design-basis margin and margins after considering increased accident loading due to the SGR/PUR.

APS Response:

Subcompartments within containment, principally the reactor cavity, the steam generator compartments, and the pressurizer compartment, are designed to withstand the transient differential pressures and jet impingement forces of a postulated pipe break. UFSAR Table 6.2-1 contains containment design parameters for postulated accidents considered for subcompartment peak pressure/temperature.

For the pressurizer subcompartment, the AOR assumes a double-ended guillotine break of the pressurizer surge line. The energy release rates were calculated for operation at PUR conditions and compared with the original plant design basis conditions. The original plant design energy release rates continue to bound the PUR plant energy release rates.

Containment Subcompartment Pressure Effects For the SG and reactor cavity subcompartments, the original design basis for concrete in the containment bounds the predicted subcompartment accident pressures determined by the RSG/PUR analysis. The analyses performed in support of PUR/RSG, which includes the results of the application of the previously approved LBB methodology to the RCS piping, reduced the maximum calculated subcompartment accident pressure loading due to elimination of the dynamic effects of RCS MCL breaks.

The original design for these subcompartments envelopes the predicted subcompartment accident pressure loading from other breaks that are postulated to occur. A separate calculation to evaluate the structural effects of the new peak accident pressure on the containment internals was not performed because it is bounded by the "original design basis.

Containment Subcompartment Temperature Effects A qualitative analysis of the accident temperature effects due to RSG/PUR on the containment internal structures was performed and compared to the original BSAP computer analysis of the internal structures. The results of this comparison determined that the original BSAP analysis remains bounding for RSG/PUR. Therefore, a detailed assessment of the design margins was not performed.

This qualitative analysis compared the RSG/PUR concrete thermal profiles that were generated using the COPATTA computer code to those values assumed in the original BSAP analysis for containment internal structures. The RSG/PUR concrete thermal profiles for the containment shell and containment basemat were used to determine the thermally induced loading and interaction forces on the containment internal structures.

The resultant loading of the subcompartment walls due to RSG/PUR was compared to the original BSAP analysis by reviewing the RSG/PUR results against the critical wall elements in the BSAP analysis that have high thermal load elements. The results of the Page 21

Mechanical & Civil Engineering Branch comparison showed that the original BSAP analysis for the subcompartment walls remains bounding.

Containment Shell Structure UFSAR Table 6.2.1-3 identifies the principal design parameters of the containment shell structure. (i.e., internal design pressure of 60 psig and design temperature of 300 OF).

These original design parameters were compared to the predicted RSG/PUR containment accident pressure/temperature profiles, which are shown in the PURLR Figures 6.2-2 and 6.2-5, respectively. The comparison determined that, although the design basis margin for containment accident pressure is decreased by RSG/PUR, the peak calculated containment accident pressures and temperatures remain bounded by the existing containment design parameters documented in UFSAR Table 6.2.1-3 (i.e.,

60 psig and 300 OF). Therefore, a separate calculation to evaluate the structural effects of the new peak accident temperature on the containment concrete was not performed.

Containment Structural Steel UFSAR Table 6.2.1-3 identifies the principal design parameters of the "Maximum DBA" temperature for Containment as 300 OF. The Main Steam Line Break (MSLB) temperature values from UFSAR Table 6.2.1-9 were compared to the SGR/PUR values from PURLR Table 6.2-6 and Figure 6.2-9 and the temperature profiles remain bounded by the existing containment design parameters.

NRC Question 12:

The acceptability of several secondary system items (i.e., steam traps) relies on an improvement in the steam quality to offset the increase in steam flow associated with PUR. Steam quality is expected to go from 0.25 percent to 0.1 percent as a result of the SGR. Clarify whether the startup testing program for PUR includes a test of the steam quality. Also, clarify whether the 0.25 percent steam quality assumed for current conditions is based on measurements or design numbers, and whether any secondary system items (i.e., steam traps) are close to their operational limits at current conditions.

APS Response:

The 0.1% moisture carryover design value stated in Section 8.8.6 of the PURLR is a RSG performance requirement. This parameter will be measured during the post-PUR startup test program.

The 0.25% moisture carryover criteria is a design value associated with the OSGs. The Unit 2 measured moisture carryover values for the OSGs are 0.6195% and 0.4840% for SG1 and SG2, respectively. These carryover values were last measured in June of 1996.

The amount of moisture carryover while operating at PUR conditions (using the anticipated 0.1% design value and the increased PUR steam mass flow rate) is less than one quarter of the actual amount of moisture carryover at current plant operating condition.

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Mechanical & Civil Engineering Branch The secondary systems capacities are not challenged at current operating levels and, based on the above discussion, should not be challenged after PUR.

NRC Question 13:

Section 9.1 of the PURLR states that a modification will be made to the main steam isolation valve bypass valve. Describe the modification.

APS Response:

The Main Steam Isolation Valve Bypass Valves (MSIVBVs) are 4 inch Anchor Darling gate valves (Model 93-15199) with Parker Hannifin air operated piston actuators (Model 14KJB2AS-13-5.250). The valves have a safety function to close on a Main Steam Isolation Signal (MSIS), and they are normally closed during full power operation of the plant.

The MSIVBVs must be modified such that they are capable of closing under the maximum, worst-case differential pressure loads. The following modifications will be made to each valve:

1. The pre-load of the actuator stanchion springs will be increased to provide increased closing thrust.

2. The air supply pressure regulator will be removed. This will allow full instrument air system pressure to be applied to the actuator to open the valve.

3. The size of the actuator air exhaust solenoid valve and exhaust piping will be increased to improve the MSIVBV's closing stroke time.

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