ML022490352

Attachment 6, Enclosure 3, Point Beach, Units 1 & 2, 1.4% Mur Power Uprate Support NRC RAI Responses Set 1 Question 4 (Non-Proprietary)

ML022490352
Person / Time
Site:	Point Beach
Issue date:	07/30/2002
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
Download: ML022490352 (135)

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RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 226 MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Attachment 6 , Point Beach Units 1 and 2, 1.4% MUR Power Uprate Support, NRC RAI Responses Set 1 Question 4 (Non-Proprietary)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ENCLOSURE 3 Point Beach Units 1 &2 1.4% MUR Power Uprate Support NRC RAI Responses Set I Question 4 Westinghouse Non-proprietary Class 3 BWestinghouse

@ 2002 Westinghouse Electric Company LLC All Rights Reserved

NSSS COMPONENTS Evaluations were performed to assess the impact on all the nuclear steam supply system (NSSS) components for the 8.7 percent power uprate for PBNP Units 1 and 2 to a core power of 1650 MWt (NSSS power of 1656 MWt). Since the NSSS design transients incorporate the effect of 60 years of operation, the results and conclusions of the component analyses are valid for an assumed 60-year operating lifetime. The component designers also reviewed the original design information related to the components. The evaluations were performed to confirm that the components continue to satisfy the applicable codes, standards, and regulatory guides under the revised conditions.

Evaluations of the following NSSS components are provided in this section.

"* Reactor Vessel and Internals

"* Control Rod Drive Mechanisms

"* Reactor Coolant Loop Piping and Supports

"* Reactor Coolant Pumps and Motors

"* Steam Generators

"* Pressurizer

• NSSS Auxiliary Equipment 1-1 07/30/02 1-1

1.0 Reactor Vessel Structural Evaluation 1.1 Introduction Evaluations were performed for various regions of the PBNP Units I and 2 reactor vessel to determine the stress and fatigue effects of the 8.7 percent core power uprate to 1650 MWt (1656 MWt NSSS power). The evaluations considered plant operation up to 60 years with renewed operating licenses.

1.2 Key Input Assumptions The key input assumptions that were needed for the power uprate are identified as the NSSS design parameters, the NSSS design transients and the reactor vessel/reactor loads internal interface loads. The temperatures and pressures considered in the reactor vessel structural evaluation are as follows:

Normal Operating Pressure: 2250 psia Normal Operating Temperatures:

Vessel Inlet Temperature Range: 526.0'F to 546.3'F Vessel Outlet Temperature Range: 592.6'F to 611.3°F Zero Load Temperature: 547.0°F Other input parameters include design inputs from the PBNP reactor vessel equipment specifications (Reference 1 for Unit I and Reference 2 for Unit 2), and revised reactor vessel and nozzle support pad loads (as necessary).

The major assumptions for the uprated conditions are as follows:

"* The vessel outlet temperature and Thor transient temperature variations apply for the internal surfaces of the outlet nozzles and closure head (including the head adapter plugs). (The main closure flanges are actually evaluated considering an intermediate temperature (Thead) that is near Thol). The vessel inlet temperature and TCoJd transient temperature variations apply for the remainder of the reactor vessel internal surfaces.

"* The design loads from References 1 and 2 (reactor vessel Equipment Specifications (E-Specs))

were considered in the reactor vessel structural analyses documented in References 3, 4 and 5, and remain conservative for the uprated NSSS design parameters, since there are no new loads of higher magnitude.

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1.3 Acceptance Criteria The acceptance criteria for the reactor vessel structural analyses and evaluations are in accordance with the applicable requirements of the 1965 Edition of Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code (Reference 6) for the Unit 1 reactor vessel and the 1968 Edition of the ASME B&PV Code with addenda through Winter 1968 (Reference 7) for the Unit 2 reactor vessel. However, the applicable acceptance criteria are the same for both reactor vessels.

The applicable acceptance criteria are described in the following paragraphs.

The maximum range of primary plus secondary stress intensity resulting from normal and upset condition design transient mechanical and thermal loads shall not exceed 3Si (Sn = design stress intensity limit) at operating temperature, in accordance with Paragraph N-414.4 of References 6 and 7. The maximum cumulative usage factor resulting from the peak stress intensities due to normal and upset condition design mechanical and thermal loads shall not exceed 1.0, in accordance with the procedure outlined in Paragraph N-415 of References 6 and 7.

The faulted conditions shall meet the component criteria of Appendix F of the ASME B&PV Code,Section III, 1974 Edition (Reference 8: The 1974 Edition was used, since Appendix F was not yet included in the 1965 and 1968 Editions). For the core support pad faulted condition analysis, the general primary membrane stress intensity limits are 2.4Sm for the Alloy 600 pads and 0.7Su (Su = ultimate tensile strength) for the low-alloy steel vessel shell. The primary membrane plus bending limits are 3.6Sm for the pads and 1.05 Su for the vessel shell.

Originally, the design requirements for the head adapter plugs conformed to the requirements for blind flanges as given in the American National Standards Institute (ANSI) B31.7 code. Reconciliation of the revised reactor coolant system (RCS) conditions with this code of construction would require no stress analysis since no structural welding is involved. However, the uprating analysis was performed in Reference 9 to comply with the reactor vessel codes of construction. Since the head adapter plug forms the primary pressure boundary along with the vessel, it is prudent to qualify the head adapter plugs to the same codes of construction as the vessels. Thus, the acceptance criteria applied in Reference 9 are the same ones previously identified for the reactor vessels from References 6 and 7. That is, the maximum range of primary plus secondary stress intensity shall not exceed 3 am and the maximum cumulative fatigue usage factor shall not exceed 1.0.

1.4 Description of Analysis/Evaluation and Results Evaluation of Design Parameters and Design Transients The NSSS design parameters affect several of the maximum ranges of stress intensity and fatigue usage factors reported in the PBNP reactor vessel stress reports (References 3, 4, and 5). Therefore, evaluations were performed for the various regions of the PBNP reactor vessels to determine the stress 1-3 07/30/02 07/30/02 1-3

and fatigue usage effects of these parameters throughout the current plant operating licenses and for up to 60 years with renewed operating licenses.

The evaluation assessed the effects of the design transients and NSSS design parameters on the most limiting locations with regard to ranges of stress intensity and fatigue usage factors in each of the regions as identified in the reactor vessel stress report and addendum (References 3, 4 and 5). The evaluation considered a worst case set of design parameters and design transients from among the high temperature/high pressure RCS conditions, the high temperature/low pressure RCS conditions, the low temperature/high pressure RCS conditions, the low temperature/low pressure RCS conditions, and the original design basis.

Where appropriate, revised maximum ranges of stress intensity and maximum usage factors were calculated. In other cases, the original design basis stress analysis remains conservative, so no new calculations were necessary, and the maximum ranges of stress intensity and fatigue usage factors reported in the Babcock and Wilcox Co. (B&W) stress report for Unit I (Reference 3) or the Combustion Engineering stress report for Unit 2 (References 4 and 5) continue to govern.

Considering any combination of the design basis and the design transients for the specified numbers of occurrences, the reactor vessel stress and fatigue analyses and evaluations justify operation with a range of vessel outlet temperature (Thot) from 592.6°F up to 611.3°F and a range of vessel inlet temperature (Tcold) from 526.0°F up to 546.3°F (Reference 10). The PBNP Unit 1 reactor vessel E-Spec (Reference 1) specifies a normal operating outlet water temperature (Th0t) of 614.5°F and a normal operating inlet temperature (T,1od) of 559.5°F, but the Unit 1 reactor vessel was originally analyzed, as documented in Reference 3, using values for Thrt of 602.0°F and TCold of 552.0°F in accordance with R. E. Ginna reactor vessel E-spec 676206, Rev. 0. The PBNP Unit 2 reactor vessel was analyzed in Reference 10 for a normal operating vessel outlet temperature (Th.t) of 610.0°F and a normal operating vessel inlet temperature (Tcold) of 552.5°F, in accordance with the PBNP Unit 2 reactor vessel E-Spec (Reference 2). The design transients incorporate the maximum Thot of 611.3°F for high temperature operation and a minimum TcoId of 525.8°F for low temperature operation. Therefore, the reactor vessel evaluation at the uprate RCS conditions, in conjunction with the reactor vessel stress reports, cover reactor operation within the bounding operating temperature ranges as shown in Section 1.2.

Such operation is shown to be acceptable in accordance with both the 1965 Edition of Section III of the ASME B&PV Code for the Unit 1 reactor vessel and the 1968 Edition of Section III of the ASME B&PV Code with addenda through the Winter of 1968 for the Unit 2 reactor vessel for the remainder of the current plant licenses and for license extension up to 60 years.

The updated maximum ranges of primary plus secondary stress intensity and maximum cumulative fatigue usage factors for the PBNP Units 1 and 2 reactor vessels accounting for the revised conditions are listed in Table 1-1. All of the cumulative fatigue usage factors remain under the 1.0 limit.

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Loss of Coolant Accident Loads In addition to the design parameters and design transients for the power uprate, a new set of loss of coolant accident (LOCA) loads at the reactor vessel/reactor internals interfaces was identified. The revised interface loads were evaluated by comparing them with the corresponding faulted condition reactor vessel/reactor internals interface loadings, which were justified for application to the PBNP Units 1 and 2 reactor vessels in Reference 10. All of the LOCA loads associated with the revised RCS conditions were found to be less than the corresponding loadings evaluated in Reference 10, and are, therefore, acceptable for application to the reactor vessel.

Seismic Analysis The reactor pressure vessel system seismic analysis is not modified as a result of the power uprate, since neither the seismic response spectra nor the mass inputs for the equipment are changed.

Reactor Vessel Head Adapter Plug Analysis A fatigue and stress evaluation of the reactor vessel head adapter plug is provided in Reference 9. This evaluation was performed to requalify the head adapter plug to new thermal transients for the revised conditions. The design complies with the requirements of Westinghouse Equipment Specification No. 953207, Revision 0, dated November 11, 1976, and the requirements of Section III of the ASME B&PV Code (1965 Edition for Unit 1 and 1968 Edition with addenda through Winter 1968 for Unit 2).

1.5 Conclusions PBNP reactor vessel operation at the NSSS power level of 1656 MWt is justified for the remainder of the current operating license and for 60 years with license renewal.

1.6 References

1. Westinghouse Equipment Specification G-676243, Rev. 0 for the Two Creeks Nuclear Station #1, WEP-105 (PBNP 1) Reactor Vessel, May 5, 1966.

2. Westinghouse Equipment Specification 676413, Rev. 1 for General Reactor Vessel, dated October 25, 1967 and Addendum Equipment Specification 677456, Rev. 2 for Wisconsin Electric Power Reactor Vessel, WIS- 105 (Point Beach 2), July 6, 1971.

3. Babcock and Wilcox Co., "Wisconsin Reactor Vessel Design Report for B&W Contract No. 610-0115-51," January 1971.

4. Combustion Engineering, Inc. Report No. CENC-1 166, "Analytical Report for Wisconsin Electric Power Reactor Vessel," October 1971.

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5. Combustion Engineering, Inc. Report No. CENC- 1166 Addendum 1, "Addendum 1 to Analytical Report for Wisconsin Electric Power Reactor Vessel," May 1972.

6. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, 1965 Edition, American Society of Mechanical Engineers, New York.

7. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, 1968 Edition with addenda through Winter 1968, American Society of Mechanical Engineers, New York.

8. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, 1974 Edition, American Society of Mechanical Engineers, New York.

9. WCAP- 14418, "Point Beach RSG Program Head Adapter Plug Analysis," June 1995.

10. WCAP-14448, "Addendum to the Stress Reports for the Point Beach Unit Nos. 1 and 2 Reactor Vessels (RSG/Uprating Evaluation)," S. L. Abbott, August 1995.

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Table 1-1 PBNP Reactor Vessel Regions Maximum Stress Intensity Ranges and Maximum Cumulative Fatigue Usage Factors Location PL+ Pb+ Q Range Uc Outlet Nozzles Nozzle: 48.8 ksi < 3 Sm= 80.1 ksi 0.122 < 1.0 Unit I Safe End: 35.8 ksi < 3 Sm = 49.2 ksi Unit 2 Safe End: 42.3 ksi < 3 Sm = 49.2 ksi Inlet Nozzles Nozzle: 40.6 ksi < 3 Sm = 80.1 ksi 0.155 < 1.0 Unit 1 Safe End: 39.6 ksi < 3 Sm= 52.9 ksi Unit 2 Safe End: 34.6 ksi < 3 S=, = 52.9 ksi Main Closure Flange Region:

1. Closure Head 70.0 ksi < 3 Sm= 80.1 ksi 0.015 < 1.0 Flange

2. Vessel Flange 67.4 ksi < 3 Sm = 80.1 ksi 0.945 < 1.0

3. Closure Studs 109.3 ksi < 3 Sm= 118.8 ksi 0.930 < 1.0 CRDM Housings 45.7 ksi < 3 Sm = 69.9 ksi 0.293 < 1.0 Vent Nozzle 25.9 ksi < 3 Sm = 69.9 ksi 0.0 < 1.0 Safety Injection Nozzles 46.8 ksi < 3 Sm = 80.1 ksi 0.200 < 1.0 External Supports 41.2 ksi < 3 Sm = 80.1 ksi 0.995 < 1.0 Brackets Vessel Wall Transition 32.2 ksi < 3 Sm = 80.1 ksi 0.004 < 1.0 Bottom Head Juncture 28.6 ksi < 3 Sm = 80.1 ksi 0.004 < 1.0 Bottom Head 57.8 ksi < 3 Sm = 69.9 ksi 0.384 < 1.0 Instrumentation Tubes Core Support Guides 57.5 ksi < 3 Sm = 80.1 ksi 0.731 < 1.0 Head Adapter Plugs 27.63 ksi < 3 Sm = 48.0 ksi 0.002 < 1.0 Note that the maximum stress intensity ranges and maximum cumulative usage factors are not marked as proprietary, as they appear in the PBNP FSAR.

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2.0 Reactor Internals Evaluations were performed to assess the impact on the reactor internal components of a power uprate for PBNP Units 1 and 2 to a core power of 1650 MWt (NSSS power of 1656 MWt) for up to 60 years with renewed operating licenses.

2.1 Introduction The reactor pressure vessel (RPV) system consists of the reactor vessel, reactor internals, and fuel and control rod drive mechanisms. The reactor internal's function is to support and orient the reactor core fuel assemblies and control rod assemblies, absorb control rod assembly dynamic loads, and transmit these and other loads to the reactor vessel. The reactor vessel internal components also function to direct coolant flow through the fuel assemblies (core), to provide adequate cooling flow to the various internals structures, and to support in-core instrumentation. They are designed to withstand forces due to structure deadweight, pre-load of fuel assemblies, control rod assembly dynamic loads, vibratory loads, and earthquake accelerations.

Operating a plant at conditions (power and temperature) other than those considered in the original design requires that the reactor vessel system/fuel interface be thoroughly addressed in order to assure compatibility, and that the structural integrity of the reactor vessel/internals/fuel system is not adversely affected. In addition, thermal-hydraulic analyses are required to determine plant specific core bypass flows, pressure drops, and upper head temperatures in order to provide input to the LOCA and non LOCA safety analyses, as well as NSSS performance evaluations.

Generally, the areas that are potentially most affected by changes in system operating conditions are:

• Reactor internals system thermal/hydraulic performance

"* Rod control cluster assembly (RCCA) scram performance

"* Reactor internals system structural response and integrity The major components and features of the reactor internals system for the PBNP units are summarized as follows. The lower core support assembly consists of the lower support plate, lower support columns, and lower core plate and core barrel, and supports the fuel assemblies on the sides and at the bottom.

The guidance and alignment of the lower core support assembly during insertion into the reactor vessel is provided by the radial support system and the head-vessel alignment pins, and special temporary guide studs attached to the vessel. The hold-down spring rests on top of the flange of the lower core support assembly. The upper core support assembly consists of the upper support plate, upper support columns, and upper core plate, and rests on top of the hold-down spring. The guidance and alignment of the upper core support assembly, during its insertion, is provided by the head-vessel alignment pins, the upper core plate alignment pins in the core barrel assembly, and the special temporary guide studs attached to the vessel. The alignment of the core, i.e., each fuel assembly, is provided through the engagement of the 07/30/02 2-1

lower core plate fuel pins into the bottom of the fuel assemblies and the upper core plate fuel pins into the top of the fuel assemblies. The vessel upper head compresses the hold-down spring, providing joint preload.

The core barrel, which is part of the lower core support assembly, provides a flow boundary for the reactor coolant. When the primary coolant enters the reactor vessel, it impinges on the side of the core barrel and is directed downward through the annulus formed by the gap between the outside diameter of the core barrel and the inside diameter of the vessel. The flow then enters the lower plenum area between the bottom of the lower support plate and the vessel bottom head and is redirected upward through the core. After passing through the core, the coolant enters the upper core support region and then proceeds radially outward through the reactor vessel outlet nozzles. The perforations in the various components, such as the lower support plate, control and meter the flow through the core.

The purpose of this section is to summarize the work performed to assess the effect on the reactor pressure vessel/internals system due to core power uprate. Also, PBNP Unit 2 performed a baffle bolt replacement program and the results of that effort are addressed in this report.

2.2 Key Input Parameters The principal input parameters utilized in the reactor internal components and RPV system are the NSSS design parameters for the power uprate. For structural analysis evaluations, the NSSS design transients and heating rates were considered.

2.3 Acceptance Criteria The main applicable criteria are described for the areas evaluated:

Thermal-Hydraulic Performance The reactor pressure vessel system is evaluated to assess the thermal-hydraulic performance which serve as inputs to other analyses including the reactor internals structural analysis, LOCA, and non-LOCA analyses. There are no acceptance criteria related to this output.

Control Rod Drop Analysis The rod drop time values generated, consistent with plant operating parameters and configuration, should be within the limit defined in the Technical Specifications. The current RCCA drop time limit is 2.2 seconds.

Seismic and LOCA Analysis of Reactor Vessel and Internals The interface loads and the time history nodal displacements of the reactor internals components, determined in the LOCA and seismic analyses, serve as inputs to various structural analyses of the internals, fuel, and the vessel. There are no acceptance criteria related to this output.

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"* Flow-Induced Vibration Response The flow-induced vibration (FIV) response of reactor internal components, in general, depends upon reactor vessel inlet flow rates (such as mechanical design flow), reactor vessel inlet temperature and reactor vessel outlet temperature. The response of the lower internals (core barrel) depends on the vessel inlet temperature and the inlet flow rates. The response of the upper internals (guide tubes and upper support columns) depends on the vessel outlet temperature and the flow exiting through the outlet nozzles. The acceptance criteria for the flow-induced vibration response are that the stresses from the FIV amplitudes remain within the endurance limit of the material for high cycle fatigue and that component loads are within acceptable limits.

"* Structural Beam Data for LOCA Forcing Functions The beam data serves as input to the generation of LOCA hydraulic forces. There are no acceptance criteria related to this output.

"* Structural Adequacy of Reactor Internal Components The PBNP internals were designed prior to the introduction of Subsection NG of the ASME B&PV Code Section IEI. Although PBNP is not committed to comply with the requirements of the Code, the design of the PBNP reactor internals was evaluated using the methods of Subsection NG of the ASME B&PV,Section III (Reference 1).

2.4 Evaluation and Results for the Power Uprate Westinghouse has performed evaluations to assess the effect of the power uprate on the reactor pressure vessel/internals system of the PBNP units.

2.4.1 Thermal-Hydraulic System Evaluations 2.4.1.1 System Pressure Losses The principal RCS flow route through the reactor pressure vessel system at the PBNP units begins at the two inlet nozzles. At this point, flow turns downward through the reactor vessel/core barrel annulus.

After passing through this downcomer region, the flow enters the lower reactor vessel dome region. This region is occupied by the internals energy absorber structure, lower support columns, bottom-mounted instrumentation columns, and supporting tie plates. From this region, flow passes upward through the lower core plate, and into the core region. After passing up through the core, the coolant flows into the upper plenum, turns, and exits the reactor vessel through the two outlet nozzles. Note that the upper plenum region is occupied by support columns and RCCA guide columns.

A key area in evaluation of core performance is the determination of hydraulic behavior of coolant flow within the reactor internals system, i.e., vessel pressure drops, core bypass flows, RPV fluid temperatures, and hydraulic lift forces. The pressure loss data is necessary input to the LOCA and 07/30/02 2-3

non-LOCA safety analyses and to overall NSSS performance calculations. The hydraulic forces are considered in the assessment of the structural integrity of the reactor internals, core clamping loads generated by the internals hold-down spring, and the stresses in the reactor vessel closure studs.

The analysis determined the distribution of pressure and flow within the reactor vessel, internals, and the reactor core. Results were obtained with a full core of Westinghouse 422VANTAGE+ fuel without intermediate flow mixing (IFM) grids, thimble plugs removed, and at RCS conditions and provided as inputs to the various engineering groups for their use.

2.4.1.2 Bypass Flow Analysis Description of Analyses Bypass flow is the total amount of reactor coolant flow bypassing the core region and is not considered effective in the core heat transfer process. Variations in the size of some of the bypass flow paths, such as gaps at the outlet nozzles and the core cavity, occur during manufacturing or change due to fuel assembly changes. Plant-specific, as-built dimensions are used in order to demonstrate that the bypass flow limits are not violated. Therefore, analyses are performed to estimate core bypass flow values to either ensure that the design bypass flow limit for the plant will not be exceeded or to determine a revised design core bypass flow.

The present design core bypass flow limit is 6.5 percent of the total reactor vessel flow with the elimination of thimble plugging devices. The purpose of this evaluation is to ensure that the design value of 6.5 percent can be maintained at the RCS conditions of the uprate. The principal core bypass flow paths are as follows:

• Baffle-Barrel Region

"* Vessel Head Cooling Spray Nozzles

"* Core Barrel - Reactor Vessel Outlet Nozzle Gap

"* Fuel Assembly - Baffle Plate Cavity Gap

"* Fuel Assembly Thimble Tubes Bypass Flow Analysis Results Fuel assembly hydraulic characteristics and system parameters, such as inlet temperature, reactor coolant pressure, and flow were used to determine the impact of uprated RCS conditions on the total core bypass flow. The design core bypass flow value of 6.5 percent of the total vessel flow can be maintained.

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2.4.1.3 Hydraulic Lift Forces An evaluation was performed to estimate hydraulic lift forces on the various reactor internal components for the uprate. This was done to ensure that the reactor internals assembly would remain seated and stable for all conditions. If the impact of the proposed changes on lift forces is found to be significant, then the estimated hydraulic lift forces would be combined with other mechanical and body forces to evaluate the resultant pre-load of the core barrel flange against the reactor vessel. Since the impact of the proposed changes on hydraulic lift forces did not show any significant effect in hydraulic lift forces compared to the present analyzed condition, it was not necessary to combine the lift forces with other forces to calculate the impact on core barrel flange pre-load. Based on the evaluation, the PBNP reactor internals will remain seated and stable for the uprated RCS conditions.

2.4.1.4 Baffle Joint Momentum Flux and Fuel Rod Stability Baffle jetting is a hydraulically induced instability or vibration of fuel rods caused by a high velocity jet of water. This jet is created by high pressure water being forced through gaps between the baffle plates which surround the core. The baffle jetting phenomenon could lead to fuel cladding damage. With the "converted upflow" baffle barrel region configuration and all baffle plate edge bolts intact and functional, the momentum flux margins remain acceptable. For Point Beach Unit 2, with the new baffle plate bolting pattern, the results of the momentum flux evaluation indicated acceptable margins of safety.

2.4.1.5 RCCA Drop Time Evaluation The RCCAs represent perhaps the most critical interface between the fuel assemblies and the other internal components. It is imperative to ensure that the uprated RCS conditions will not adversely impact the operation of the RCCAs, either during accident conditions or normal operation.

The evaluation determined the potential impact of the uprated conditions on the RCCA drop time. The results indicate that the current maximum RCCA drop time of 2.2 seconds (which includes the seismic allowance) remains applicable for the accident analysis.

2.4.2 Mechanical System Evaluations Changes in RCS conditions generally impact the performance of the reactor pressure vessel and its internals under all modes of operation. It is, therefore, important that the mechanical response of the reactor pressure vessel and its internals be evaluated. The mechanical system evaluations that are affected due to changes in the RCS conditions in general, consist of the following:

• Response due to LOCA

• Response due to flow-induced vibrations 2-5 07130/02 07/30/02 2-5

2.4.2.1 LOCA Evaluations Descriptions of Analysis To perform the RPV LOCA analysis of the PBNP units, a finite element model of the RPV system is developed.

The mathematical model of the RPV is a three-dimensional nonlinear finite element model that represents the dynamic characteristics of the reactor vessel and its internals in the six geometric degrees of freedom. The model was developed using the WECAN (Reference 2) computer code. The WECAN computer code (or predecessor codes) has been used for this analysis since the original plant design.

In order to evaluate the impact of changes in RCS conditions on the dynamic response of the RPV system, LOCA analyses were performed to generate core plate motions and the reactor vessel/internals interface loads. The core plate motions are then used to evaluate the structural integrity of the core.

Since application of leak-before-break (LBB) methodology has been licensed for the main coolant loop, consideration of breaks in the main coolant loop are not required for structural evaluations. The next limiting breaks to be considered are the branch line breaks, such as:

"* Accumulator line

"* Pressurizer surge line

"* Residual heat removal line The most limiting breaks considered for the dynamic analysis are:

"* Accumulator line (cold leg) and

"* Pressurizer surge line (hot leg) breaks The hydraulic LOCA forces for accumulator line break and pressurizer surge line break are used in the reactor vessel LOCA analysis.

Following a postulated LOCA, forces are imposed on the reactor vessel and its internals. These forces result from the release of the pressurized primary system coolant and, for auxiliary pipe breaks, from the disturbance of the mechanical equilibrium in the piping system prior to the rupture. The release of pressurized coolant results in traveling depressurization waves in the primary system. These depressurization waves are characterized by a wavefront with low pressure on one side and high pressure on the other. The wavefront translates and reflects throughout the primary system until the system is completely depressurized. The rapid depressurization results in transient hydraulic loads on the mechanical equipment of the system.

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The LOCA loads applied to the reactor pressure vessel system consist of (1) reactor internal hydraulic loads (vertical and horizontal), and (2) reactor coolant loop mechanical loads. All the loads are calculated individually and combined in a time history manner.

The MULTIFLEX computer code (Reference 3) calculates the hydraulic transients within the entire primary coolant system. It considers sub-cooled, transition, and two-phase (saturated) blowdown regimes. The MULTIFLEX program employs the method of characteristics to solve the conservation laws, and assumes one-dimensionality of flow and homogeneity of the liquid-vapor mixture.

Results of Reactor Pressure Vessel LOCA Analysis The severity of a postulated break in a reactor vessel is related to two factors: the distance from the reactor vessel to the break location and the break opening area. The nature of the reactor vessel decompression following a LOCA, as controlled by the internals structural configuration previously discussed, results in larger reactor internal hydraulic forces for pipe breaks in the cold leg than in the hot leg (for breaks of similar area and distance from the RPV). Pipe breaks farther away are less severe because the pressure wave attenuates as it propagates toward the reactor vessel. Therefore, pipe breaks at the reactor vessel inlet nozzle are more severe, because of the absence of pressure wave attenuation and the structural configuration of the core. In general, the auxiliary line breaks like accumulator line and the pressurizer surge line breaks are not as severe as the main line breaks such as RPV inlet nozzle or reactor coolant pump outlet nozzle break.

The results of reactor vessel displacements and the impact forces calculated at vessel/internals interfaces are used to evaluate the structural integrity of the reactor vessel and its internals. The core plate motions for both breaks were used in fuel grid crush analysis and to confirm the structural integrity of the fuel.

The results of the reactor internals analyses demonstrate continued compliance with applicable codes, standards, and regulatory criteria at the uprated RCS conditions.

Subsequent to the RSG project, analyses were performed for specific baffle bolting patterns for PBNP Unit 2. The results indicate that for the Point Beach Unit 2 configuration (i.e., converted upflow) the maximum grid impact loads for these faulted conditions are relatively insensitive to the bolting pattern for small variations around a basic pattern.

2.4.2.2 Flow-Induced Vibrations Flow-induced vibrations of pressurized water reactor internals have been studied at Westinghouse for a number of years. The objective of these studies is to assure the structural integrity and reliability of reactor internal components. These efforts have included in-plant tests, scale-model tests, as well as tests in fabricators' shops and bench tests of components, along with various analytical investigations. The results of these scale-model and in-plant tests indicate that the vibrational behavior of two-, three-, and four-loop plants is essentially similar, and the results obtained from each of the tests compliment one another and make possible a better understanding of the flow-induced vibration phenomena.

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The flow-induced vibration amplitudes and stresses are in general very small and well below the ASME code allowable fatigue curve. Note also that the ASME Code (Reference 4) allowable fatigue curve is extremely conservative. Based on the results of the analysis detailed in Reference 5, the structural integrity of the reactor internals remain acceptable with regards to flow-induced vibrations and are valid for 60 years given that the critical interface gaps do not significantly degrade. Inspection of the critical interface gaps (e.g., lower radial key and upper core plate alignment pin) must be addressed during the license renewal process.

Subsequent to the SGR project, baffle-barrel region bolting analyses were performed for specific baffle bolting patterns for PBNP Unit 2. Bolt fatigue stresses were calculated by harmonic analysis for one beam mode and two shell modes of the core barrel vibration. The fatigue stresses calculated were then combined by SRSS and compared to the fatigue endurance limit of 316 cold-worked (CW) stainless steel.

In all cases calculated, the resultant fatigue stresses were found to be below the bolt material endurance limits. Alternating stresses, SaIt, were determined to be well below the ASME code allowable for all cases.

For Point Beach Unit 2, the new replacement baffle-former bolts are of a Non-Westinghouse design. The assumption made in the baffle-former bolt replacement analyses, performed by Westinghouse, was that the replacement bolt has the geometric, thermal, dynamic and structural properties equal to the Westinghouse replacement baffle bolt. For Unit 2 the specific baffle bolt replacement analysis the design life was set to 40 years. This would be 40 years from the time of replacement. Therefore, the design life of the new baffle bolts in Unit 2 would be extended to year 64 in plant life.

2.4.3 Structural Evaluation of Reactor Internal Components In addition to supporting the core, a secondary function of the reactor vessel internals assembly is to direct coolant flows within the vessel. While directing the primary flow through the core, the internals assembly also establishes secondary coolant flow paths for cooling the upper regions of the reactor vessel and for cooling the internals structural components. Some of the parameters influencing the mechanical design of the internals lower assembly are the pressure and temperature differentials across its component parts and the flow rate required to remove the heat generated within the structural components due to radiation (e.g., gamma heating). The configuration of the internals provides for adequate cooling capability.

Structural evaluations are performed to demonstrate that the structural integrity of the reactor components is not adversely affected directly by the change in RCS conditions and transients and/or by secondary effects of the change on reactor thermal-hydraulic or structural performance. The presence of heat generated in reactor internal components, along with the various fluid temperatures, results in thermal gradients within and between components. These thermal gradients result in thermal stresses and thermal growth, which must be accounted for in the design and analysis of the various components.

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2.4.3.1 Component Evaluation This section summarizes the results of structural evaluations performed for the reactor internals components judged to be most key for the uprated RCS conditions. Westinghouse performed a review and an evaluation of the effects of the NSSS design transients and the power uprate on the following key reactor internal components:

"* Lower Core Plate

"* Lower Core Support Plate

"* Lower Support Columns

"* Core Barrel Outlet Nozzle Projection

"* Core Barrel Flange

"* Lower Radial Restraints (Clevis Insert)

• Upper Core Plate Alignment Pin

"* Upper Support Columns

"* Upper Core Support Plate

"* Upper Core Plate

• Guide Tubes and Support Pins

• Flexureless Insert The results of this structural evaluation indicate that the change in RCS thermal transients due to uprated RCS conditions does not significantly affect the stress and fatigue usage factors. The margins of safety and cumulative fatigue usage factors for some of the reactor internals components are provided in Table 2-1.

Baffle-Barrel Region Evaluations For the baffle-barrel region evaluation, the heat generation rates used are based on the availability of hafnium power suppression rods.

The baffle-barrel regions consist of a core barrel into which baffle plates are installed, supported by bolting interconnecting former plates to the baffle and core barrel. The baffle-to-former bolts restrain the motion of the baffle plates that surround the core. These bolts are subjected to primary loads consisting of deadweight, hydraulic pressure differentials, seismic loads, as well as secondary loads consisting of preload and thermal loads resulting from RCS temperatures and gamma heating rates. The power uprate program does not effect the deadweight or seismic loads. The baffle-to-former bolt thermal loads are induced by differences in the average metal temperature between the core barrel and baffle plate.

07/30/02 2-9

It is assumed that for PBNP Unit I all baffle-former bolts are intact and for PBNP Unit 2 all replacement baffle-former bolts are intact. For PBNP Unit 2, the baffle-former bolt qualification is based on the assumption that the replacement bolt has structural properties that result in stresses that are equal to or less than stresses in the original designed bolt for a given load. The basis used to determine the acceptability of the baffle-former bolts are previous evaluations performed for the PBNP units as part of the Point Beach upflow conversion program.

The gamma heating rates seen by the baffle-barrel region would increase proportionally with the increase in power. The effect of this increase in gamma heating rates was evaluated as follows. For PBNP Unit 1, the difference in temperature between the baffle and core barrel for the core power of 1650 MWt (with hafnium absorbers) was less than that from previous evaluations, so the baffle-former bolt displacements remain acceptable. For PBNP Unit 2 (with hafnium absorbers intact), the baffle bolt displacement seen in the remaining bolts was less than the worst bolt for the nominal, all-bolts-intact pattern.

The maximum equivalent linear temperature difference (ATbaf) for the baffle plate was less than the previous evaluation and the baffle plate remains acceptable.

No new cumulative usage factor (CUF) calculations were performed for the baffle-barrel region components since it has been shown that previous evaluations performed for the PBNP units remain bounding.

Lower Core Plate Structural Analysis The lower core plate structural analyses use the heat generation rates which assume that the hafnium absorbers have been removed.

The lower core plate is a perforated circular plate that supports and positions the fuel assemblies. The plate contains numerous holes to allow fluid flow through the plate. The fluid flow is provided to each fuel assembly and the baffle-barrel region. The plate is bolted at the periphery to a ring welded to the inside diameter of the core barrel. The center span of the plate is supported by the lower support columns, which are attached at the lower end to the lower support plate.

A structural evaluation was performed to demonstrate that the uprate does not adversely affect the structural integrity of the lower core plate. The uprate causes an increase in the heat generated within the lower core plate. The conclusion of the evaluation is that the structural integrity of the lower core plate is maintained.

Upper Core Plate Structural Analysis The upper core plate structural analyses use the heat generation rates which assume that the hafnium absorbers have been removed.

2-10 07/30/02 07/30/02 2-10

The upper core plate positions the upper ends of the fuel assemblies and the lower ends of the control rod guide tubes, thus serving as the transitioning member for the control rods in entry and retraction from the fuel assemblies. It also controls coolant flow in its exit from the fuel assemblies and serves as a boundary between the core and the exit plenum. The upper core plate is restrained from vertical movement by the upper support columns, which are attached to the upper support plate assembly. Four equally spaced core plate alignment pins restrain the lateral movement.

A structural evaluation was performed to demonstrate that the uprate does not adversely affect the structural integrity of the upper core plate. The uprate causes an increase in the heat generation within the upper core plate. The conclusion of the evaluation is that the structural integrity of the upper core plate is maintained.

2.6 Conclusions Evaluations have been performed to assess the effect of the uprate on the reactor pressure vessel/internals system. The results of these analyses are summarized below:

1. The vessel pressure losses, bypass flows, hydraulic lift forces, and baffle joint momentum flux/fuel stability are acceptable for the uprate.

2. The design core bypass flow value of 6.5 percent of the total vessel flow can be maintained.

3. The current RCCA drop time Technical Specification limit of 2.2 seconds remains valid.

4. The structural integrity of the reactor internals is maintained with the uprate.

The evaluations performed for the PBNP units for the power uprate continue to comply with the current PBNP licensing basis acceptance requirements. Acceptable margins of safety and fatigue utilization factors have been demonstrated. All evaluations documented in this section utilized the NSSS design transients which have been validated for a postulated 60-year plant life. The margins of safety and cumulative fatigue usage factors that resulted from the evaluations are shown in Table 2-1.

2.7 References

1. ASME Boiler and Pressure Vessel Code, Section III-NG, 1989 Edition.

2. "WECAN/Plus Users' Manual," Second Edition, (Westinghouse Proprietary), December 1, 1992.

3. WCAP-8708-P/A, "MULTIFLEX, a FORTRAN-IV Computer Program for Analyzing Thermal Hydraulic Structure System Dynamics," (Westinghouse Proprietary), September 1977.

4. ASME Boiler and Pressure Vessel Code, Section 1I1, Appendices 1989 Edition.

08/01/02 2-11

5. WCAP-14459, "Reactor Pressure Vessel and Internals System Evaluations for The Point Beach Units 1 and 2 Power Uprating/Replacement Steam Generator Program," April 1996.

2-12 08/01/02 2-12

Table 2-1 PBNP Units 1 and 2 Summary of Safety Margins and Cumulative Usage Factors Margin of Safety (1) Cumulative Usage Factor Reactor Internal Component Pm+Pb+Qm+Qb (CUF)

[ bc [ b,c Lower Core Support Plate

] b,c b,c Lower Support Columns [

Core Barrel Flange [ bc r ] bc b,c Guide Tubes and Support Pins [ bc

[ ] b,c [ bc Lower Core Plate Baffle-Barrel Region See Note 2 See Note 2

[ bc ] b,c Upper Core Plate Note:

1. Margin of Safety = (Allowable/Actual) - 1

2. No new cumulative usage factor (CUF) calculations were performed for the baffle-barrel region components since it has been shown that previous evaluations performed for the PBNP units remain bounding.

Bracketed [ ] b,c information designates data that is Westinghouse Proprietary.

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3.0 Control Rod Drive Mechanisms Evaluations were performed to assess the impact on the control rod drive mechanisms (CRDMs) of a power uprate for PBNP Units I and 2 to a core power of 1650 MWt for up to 60 years with renewed operating licenses.

3.1 Introduction This section addresses the acceptability of the Westinghouse CRDMs based on the uprate NSSS design parameters. Both of the PBNP units have L-106A CRDMs, full-length (F/L) mechanisms manufactured by the Westinghouse Electro-Mechanical Division (WEMD). The original design of the PBNP units included part-length (P/L) mechanisms. The P/L CRDMS have been removed from Unit 1, however, they are still in place in Unit 2. The P/L CRDMs were manufactured by Royal Industries. This section addresses the ASME Code (Reference 1) aspects of the revised parameters/transients for the F/L and P/L CRDMs.

3.2 Key Input Assumptions The NSSS design parameters and the NSSS design transients were used to evaluate the impact of the uprate on the CRDM.

3.3 Acceptance Criteria The purpose of this evaluation is to demonstrate the continued applicability of the original E-Specs (References 2 and 3) and generic CRDM Code evaluations (References 4, 5 and 6) in order to demonstrate compliance with applicable requirements of the ASME Code (Reference 1). The critical components that make up the ASME Code pressure boundary covered by this evaluation are the head adapter, the latch housing, the rod travel housing, the cap, and the canopies provided for the seal welds at the lower, middle and upper joints. These components were evaluated for the F/L CRDMs in Reference 4 for both steady-state and transient conditions as required by Section III, Subsection NB, of the ASME Code. Compliance with specific ASME Code requirements for pressure boundary thickness, material stresses and stress intensities, and fatigue usage factors is shown in References 4 and 5. This evaluation will demonstrate continued compliance with these requirements. The Code stress analyses for the P/L CRDMS are given by the Royal Industries reports of Reference 6, performed to the P/L CRDM E-Spec (Reference 3).

3.4 Description of Analysis/Evaluation and Results The design parameters used for the CRDM evaluation are the hot-leg temperature data, which is the vessel outlet temperature. The temperature increase from 602.8°F (at 1518.5 MWt core power) to the bounding temperature of 611.3°F is only 8.5°F. For the full-length and part-length CRDMs, the conditions in the generic analyses (References 4, 5 and 6) bound the NSSS design parameters for the 07/30/02 3-1

power uprate, since the generic analyses used 650'F as the hot leg maximum temperature versus 611.3°F used for the power uprate evaluation.

For the part-length CRDMs, the Code analysis is given by Reference 6. Part-length CRDMs are no longer actively used; however, the housings remain on Unit 2. An evaluation of the P/L CRDM was performed which concluded that the stress analysis of the full-length CRDM lower joint may be used as a basis to justify the part-length CRDM lower joint. In addition, since thermal transients do not significantly affect the P/L CRDM above the thrust bearing retainer assembly, no additional analysis will be required to justify the design transients for the part-length CRDM.

It should be noted that for the F/L CRDMs, most system transients do not fully affect the pressure boundary parts since the transients are of relatively short duration and internal fluid convection is low.

However, the generic analysis conservatively assumes that the primary fluid temperature change is experienced by the CRDM wetted surfaces.

The NSSS design transients applicable to the uprate were compared to the transients used in the full length E-Spec and the generic analysis. The E-Spec or the generic analysis values bound the power uprate values except for the transients that are discussed in Section 3.4.1 (Large Step-Load Decrease, Loss of Load, Loss of Flow, Reactor Trip). The power uprate RCS conditions also require consideration of some transients (Feedwater Cycling, Boron Concentration Equalization, Loss of Power, Inadvertent Actuation of Auxiliary Spray, Steam Line Break, and Turbine Roll Tests) that were not previously considered applicable to the PBNP CRDMs. These "new" transients were not previously considered applicable to the PBNP Units 1 and 2 CRDMs because they were not defined in the original equipment specification for the PBNP CRDMs. These new transients are discussed in Section 3.4.2.

Also, the present NSSS design transients bound 60 years of operation except for the hydro-pressure test cycles. An evaluation of the additional cycles is presented in Section 3.4.3.

3.4.1 Large Step-Load Decrease, Loss of Load, Loss of Flow and Reactor Trip The large step-load decrease carries a total pressure fluctuation AP = 482 psi. This does not affect the fatigue waiver/analysis since it is not a "significant pressure fluctuation" per Reference 1. Per N-415.1 (b), a significant pressure fluctuation is one that exceeds the following quantity:

1 Sa at 106 cycles Significant pressure fluctuation, AP = Design Pressure x - x 3 sin 3-2 07/30/02 07/30/02 3-2

where:

S - allowable amplitude of alternating stress intensity from the design fatigue curve given in the ASME Code for the material under consideration. For 304SS at 550'F, Sa at 106 cycles = 26 ksi.

Sm allowable design stress intensity for the material under consideration from Table N-421, N-422, or N-423 of the ASME Code. For 304SS at 550'F, Sm = 16.9 ksi.

Significant pressure fluctuation, AP = 2500 (1-j 21269 = 1282 psi Thus, the transient with a AP = 482 psi is not considered a significant pressure fluctuation to be included in a fatigue waiver analysis.

The pressure fluctuations for the remaining transients are, Loss of Load AP = 397.8 psi, Loss of Flow AP = 313 psi and the Reactor Trip AP = 455 psi. These transients are also not considered to be significant pressure fluctuations and do not affect the fatigue waiver analysis.

3.4.2 New NSSS Design Transients The NSSS design transients that were not previously applicable, but which now must be considered for the PBNP CRDMs are: Feedwater Cycling, Boron Concentration Equalization, Loss of Power, Inadvertent Actuation of Auxiliary Spray, Steamline Break and Turbine Roll Tests.

The Steamline Break does not affect the fatigue since it is a faulted case. Also, since the pressure decreases, the primary stresses decrease.

These transients have pressure drops (or APs) of less than 1282 psi and, hence, do not affect the pressure transient fatigue waiver analysis. The Feedwater Cycling transient has a large number (25,000) of temperature fluctuation cycles of AT = 32°F. Again per N-415.1 (d), a significant temperature fluctuation, AT is:

Significant temperature fluctuation, AT = Sa 2Ea where:

Sa allowable amplitude of alternating stress intensity from the design fatigue curve given in the ASME Code for the material under consideration. For 304SS at 550'F, Sa at 106 cycles = 26,000 psi.

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E modulus of elasticity for the material under consideration from Table N-427 of the ASME Code. For 304SS at 550TF, E = 25.8 x 106 psi.

a = the instantaneous coefficient of thermal expansion of the material under consideration at the mean temperature under consideration. For 304SS at 31 0°F, a = 9.76 x 10-6 in./in./0 F.

26000 Significant temperature fluctuation, AT =

2(25.8)(9.76)

= 51.6"F Hence, the 32'F cycles are not significant and do not affect the fatigue waiver/analysis. The Loss of Power has a 42°F AT, which is also less than 51.6°F. The Turbine Roll Test has a pressure/temperature decrease, which was shown to be acceptable.

Thus, the new transients are acceptable to the CRDMs for the revised RCS conditions. Note these transients were analyzed in more detail in newer, but similar, CRDM designs and were found to meet ASME B&PV Code limits.

3.4.3 Hydrotest Cases There was an increase in the number of 2500 psia hydrotest cases from 40 (Reference 2) to the new value of primary side hydrotests of 100 (to 400'F, 2500 psig) plus 30 (90'F, 2250 psig) or 130 total. The generic analysis used the ASME B&PV Code waiver analysis and also showed that a simple, but conservative fatigue evaluation met ASME B&PV Code limits; however, hydrotest was not included.

Hence, an increase in the hydrotest cycles requires a separate evaluation. Since the membrane and the membrane plus bending primary stress intensity limits are met, only the effect on fatigue of the increased test cycles needs to be addressed.

Using the hydrotest primary membrane plus bending stress intensities (M+B S.I.) and conservatively using the ASME B&PV Code maximum fatigue reduction factor of 5, the fatigue usage contribution of the hydrotest cycles can be approximated, as shown in Table 3-1.

Inspection of the hydrotest contributory fatigue usage factors shows that essentially no fatigue damage results from the 2500 psi hydrotests. Hence, the increase in 2500 psi hydrotest cycles is acceptable for the CRDMs. The full pressure fatigue waiver criterion (Reference 5) is still met.

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3.5 Conclusion From the preceding evaluations and the previous generic analyses, it can be concluded that the design of the part-length and full-length CRDMs meets applicable ASME Code requirements at a core power rating up to 1650 MWt (1656 MWt NSSS power) for up to 60 years of operation.

3.6 References

1. ASME Boiler and Pressure Vessel Code, Section Ir-NB, Summer 1966 through Winter 1969 Addenda.

2. CRDM Equipment Specification 676426, Rev. 1, Westinghouse Atomic Power Division, November 3, 1967, and Interim Change No. 1, dated December 10, 1976, per P.A. No. 6, May 14, 1970.

3. Equipment Specification 677052, Rev. 2, "Drive Mechanism for Part Length Control Rods",

Westinghouse Atomic Power Division, September 30, 1968.

4. Engineering Memorandum 4531, Rev. 2, "Stress and Thermal Report of Type L106A and L106BControl Rod Drive Mechanism Pressure Containing Components," Westinghouse Electro-Mechanical Division, S. Ganguly, J. Raymond, and A. Reed, April 12, 1976.

5. Addendum I to Engineering Memorandum 4531, Rev. 2, "Stress and Thermal Report of Type LI06A and L106B CRDM," Westinghouse Electro-Mechanical Division, D. Brondyke, October 15, 1976.

6. Report No. 121X135, "Stress Report - Pressure Vessel Portion of 121J001 Series Control Rod Drive Mechanism," Royal Industries, by C. Hiers, February 26, 1969; and Addendum VII, May 10, 1973; and Seismic Data Report 121X141, June 19, 1969.

3-5 07130/02 07/30/02 3-5

Table 3-1 Fatigue Usage Contribution of the Hydrotest Cycles Component Fatigue Usage(1 )

Upper Joint - Cap [ ]b~c

- Rod Travel Hsg [ ]bc

- Canopy [ ]bxc Middle Joint - Rod T.H. [ ]b.c

- Latch Hsg [ ]bc

- Canopy [ ]bsc Lower Joint - Latch Hsg I ]b,c

- Head Adapter [ ]bc

- Canopy Ib,c Notes:

1. Fatigue usage factor is the 130 hydro cycles divided by the allowed cycles.

Bracketed [ ] b,c information designates data that is Westinghouse Proprietary.

07/30/02 3-6

4.0 Reactor Coolant System Piping 4.1 Introduction Evaluations were performed to assess the impact on the reactor coolant system (RCS) piping of a 8.7-percent power uprate for PBNP Units 1 and 2 to a core power level of 1650 MWt for up to 60 years with renewed operating licenses.

The NSSS design parameters for the uprating were reviewed for potential impact on the following components:

• Reactor coolant loop (RCL) piping

• Primary equipment nozzles

• Pressurizer surge line piping These components were evaluated against Reference 1 and Reference 2 allowables, or against equipment nozzle allowables (References 3 and 4) as appropriate. The following sections describe the analysis performed and summarize the results.

4.2 Key Input Parameters Three basic sets of input parameters are used in the evaluation of the components defined above.

"* NSSS design parameters

"* Thermal design transients

"* Loss of coolant accident (LOCA) forces and reactor pressure vessel LOCA displacements 4.3 Acceptance Criteria The acceptance criteria for the PBNP RCL piping stress evaluation are contained in the USAS B31.1 Power Piping Code (Reference 1), which does not require a formal fatigue analysis.

The acceptance criteria used for the piping qualification of the PBNP surge line stratification analysis are defined in Reference 5. They are based on the 1986 edition of the ASME B&PV Code Section Ill requirements (Reference 2).

4-1 07/30/02 07/30/02 4-1

4.4 Description of Analysis/Evaluation and Results 4.4.1 Deadweight and Thermal Analysis Computer structural analyses were performed on the RCL piping system model for the loading conditions of deadweight and thermal based on hot full power (HFP) conditions. The WESTDYN computer code was used for this analysis.

The thermal expansion analysis was run to give the range of loadings associated with the temperature conditions defined in the NSSS design parameter list. Two thermal expansion runs were made. The first run represents the low temperature case in which there is no lateral restraint from the primary equipment supports. As the temperature in the loop increases up to the plant's existing (prior to uprate) normal operating hot and cold leg temperatures, the primary equipment lateral supports are inactive. At these normal operating hot and cold leg temperatures, there is a nominal zero gap between the equipment and the support structure. An additional thermal expansion or a dynamic event such as a seismic occurrence will activate the lateral supports. The second thermal expansion run represents the postulated high temperature case (still normal full power) and the analysis has active lateral supports (Steam generator lower backside bumpers) that restrain the last 8.5°F thermal expansion that is above the existing normal operating hot leg temperature.

The results from both of these thermal runs were used in subsequent evaluations for the loop piping and primary equipment nozzles discussed later in this section. Results from this analysis were also used for the NSSS equipment supports evaluation (Section 9) and leak-before-break (LBB) evaluation (Section 10).

4.4.2 Seismic Analysis The loop seismic analysis was not repeated for the power uprate, since there were no changes in mass of the system. The seismic results were taken from the PBNP Unit 1 analysis that was performed in 1983 to reconcile the Unit 1 replacement steam generator. This analysis used the WESTDYN computer code.

These Unit 1 seismic results apply without modification to Unit 2 because of the similarities in the steam generators between Units 1 and 2.

Various seismic cases (from the Unit 1 analysis) were considered for the RCL analyses. All of the cases assumed that the steam generator upper support was active in both directions, and the reactor coolant pump (RCP) tie rods were assumed to be inactive in all of the cases. The steam generator lower lateral support was assumed to be either active or inactive depending on the analysis case being run.

The seismic analysis results for this RCL model were used as input to the specific evaluations for the loop piping, the primary equipment nozzles discussed later in this section. Results from this analysis were also used for the equipment supports evaluation (Section 9) and leak-before-break (LBB) evaluation (Section 10).

07/30/02 4-2

4.4.3 LOCA Analysis The loop LOCA analysis considers two different inputs - the loop LOCA forces associated with defined postulated breaks, and reactor vessel dynamic LOCA displacements associated with defined postulated break cases. The design basis for the PBNP RCL piping LOCA analysis has credited the use of loop LBB methodology. Postulated guillotine breaks in the primary loop piping have been replaced with postulated guillotine breaks at the loop branch connections for the largest class 1 auxiliary lines. LOCA forces were evaluated for three nozzle breaks, the surge line nozzle on the hot leg, the accumulator line nozzle on the cold leg, and the residual heat removal (RHR) line nozzle on the hot leg. The RHR line nozzle break conservatively uses data from the surge line break, but applies the break forces consistent with the RHR line nozzle.

4.4.4 Reactor Coolant Loop Piping The evaluation for the primary loop piping involved calculating the piping stress at the critical components and comparing them to the allowable stresses calculated from the requirements of Reference 1. Stresses for each loading are required to be combined as defined in the B3 1.1 piping analysis code, which includes deadweight, seismic, LOCA, and thermal stresses as necessary. The results are summarized in Table 4-1 and show that all piping stresses are less than allowable.

4.4.5 Primary Equipment Nozzles The evaluation for the primary equipment nozzles involved a comparison of the loads generated at uprate conditions for the thermal and LOCA load conditions with the allowable nozzle loadings for that equipment. Since seismic and deadweight loads did not change for the uprate conditions, comparisons of these loads to allowables were not required. The change in nozzle loads was considered insignificant for all but the LOCA load case. The evaluation shows that all results are within allowable limits for primary loop nozzles.

4.4.6 Pressurizer Surge Line Stratification The impact of changes in the revised NSSS design parameters, thermal design transients, and the 60-year life extension were factored into determining the ASME stress levels and allowables for the surge line.

The evaluation included a review of the fatigue analysis and the stratification loadings that were transmitted to the pressurizer nozzle from the surge line piping. The changes and the percent increases for the thermal design transients were tabulated and the impact on the fatigue usage factor was calculated. The forces and moments that were generated by the stratified conditions in the surge line also exist at the pressurizer nozzle. The power uprate conditions were reviewed to determine if the previous enveloping loads on the nozzle changed significantly. Temperature differences between the hot leg and pressurizer were used to calculate stratified moments in the surge line piping. The difference between 4-3 07/30/02 4-3

the previous Thot (hot leg temperature) and the new Thbo was determined and used in the determination of new nozzle loads.

The results of the evaluation for the pressurizer surge line stratification showed that the power uprate conditions changed the fatigue usage factor at the location of highest usage factor by a negligible amount.

The calculated change in loadings on the pressurizer nozzle due to stratification for the power uprate conditions was not considered significant and was not evaluated further. The results presented in the Reference 5 report remain unchanged. Results of the evaluation for the surge line from Reference 5 are summarized in Table 4-2.

4.5 Conclusions The revised NSSS design parameters associated with the power uprate conditions were evaluated for impact on the RCL piping, the primary equipment nozzles, and the pressurizer surge line. The evaluation indicates that all components meet appropriate allowables. The evaluation for the stated components concludes that there is no adverse effect on the ability of these components to operate for 60 years.

4.6 References

1. USA Standards B3 1.1 Power Piping Code.

2. ASME Boiler and Pressure Vessel Code, Section 11I, 1986 Edition.

3. Design Specification 955381, "Wisconsin Electric Power Company Point Beach Nuclear Plant,"

December 18, 1981. (Unit 1 Steam Generator nozzle allowables)

4. Design Specification 412A72, "Wisconsin Electric Power Company Point Beach Nuclear Plant Unit 2," November 22, 1996. (Unit 2 Steam Generator nozzle allowables)

5. WCAP-13509, "Structural Evaluation of the Point Beach Units 1&2 Pressurizer Surge Lines, Considering The effects of Thermal Stratification," October 1992.

4-4 07/30/02 4-4

Table 4-1 Reactor Coolant Loop Piping Stress Analysis Summary Hot Leg Crossover Leg Cold Leg Evaluation Maximum Allowable Maximum Allowable Maximum Allowable Equation 11 (ksi) (DW, P) [ ]bIc 15.9 [ b,C 15.9 [ jbc 15.9 Equation 12 (ksi) [ ]bC 19.08 [ ]bC 19.08 i ]bC 19.08 (DW, P, DBE)

(DW, P, DBE, LOCA) [ ]bC 19.08 [ ]bC 19.08 [ ]bC 19.08 Thermal Stress (ksi) [ ]b, C 23.85 [ ]b, C 23.85 [ ]b,C 23.85 Bracketed [ bc information designates data that is Westinghouse proprietary.

Table 4-2 Pressurizer Surge Line Piping Fatigue"1 ) Analysis Summary Surge Line ASME Evaluation Maximum Allowable Usage Factor [ ]bc 1.0 Equation 12 Stress [ ]b,c 57.9 Equation 13 Stress [ ]b,c 55.8 Note:

1. Primary Stress Evaluation was performed by Sargent & Lundy Engineers.

Bracketed [ ] b,c information designates data that is Westinghouse proprietary.

4-5 07/30/02 07/30/02 4-5

5.0 Reactor Coolant Pumps and Motors Evaluations were performed to assess the impact on the reactor coolant pumps (RCPs) and motors of a power uprate for PBNP Units 1 and 2 to a core power of 1650 MWt for up to 60 years with renewed operating licenses.

5.1 Introduction Each PBNP reactor coolant loop contains a Model 93 single-stage shaft-seal pump driven by an air cooled motor. The pump is a vertical assembly consisting of (from top to bottom) the motor, the motor support stand, the seal assembly, and the hydraulic unit.

The reactor coolant pumps and RCP motors were evaluated to determine the impact of the uprate RCS conditions. For the RCPs, the intent was to demonstrate that the RCP structural integrity is not adversely impacted by the uprate NSSS design parameters, and that this pressure boundary component continues to comply with industry codes and standards. Note that the PBNP RCPs are pre-code stamping and, hence, there are no specific plant ASME Code reports.

5.2 Key Input Assumptions The NSSS design parameters and the NSSS design transients were used to evaluate the impact of the uprate on the RCPs.

5.3 Acceptance Criteria Structural The purpose of this evaluation is to demonstrate the continued applicability of the original Equipment Specification (E-Spec) (Reference 1) and generic evaluation (References 2-6) in order to demonstrate compliance with applicable requirements of industry codes and standards. The critical components that are covered by this evaluation are the pump casing and feet, the closure flange and bolts, the thermal barrier heat exchanger assembly, and the seal housing and bolts. These components are evaluated for both steady state and transient conditions. Compliance with ASME Code (Reference 7) requirements is used to demonstrate acceptability, even though the RCPs pre-date the inclusion of pumps into the ASME Code.

Electrical The acceptability of the RCP motors for the loadings associated with the power uprate conditions is based on the loading remaining within the original nameplate ratings for the motors.

5-1 07/30/02 07/30/02 5-1

5.4 Description of Analysis/Evaluation and Results 5.4.1 Reactor Coolant Pump The RCP is located in the cold leg, downstream of the steam generator outlet. The pump normal operating pressure per the original E-Spec (Reference 1) is 2250 psia and the normal operating temperature is 550'F. The original RCP analysis was based on these values. The uprated RCS operating pressure is 2250 psia and the RCP temperature (reactor pressure vessel inlet) are the same or less than the E-Spec values. Hence, there is no effect on the RCP analyses due to the uprated NSSS design parameters.

The RCP evaluation also considered the NSSS design transients which reflect the uprate conditions and support plant license extension to 60 years. The present 40-year cycle count also bounds 60 years, except for hydro-pressure cycles. The new hydro cycle count value (130 total) is less than the E-Spec cycles of 250 hydros at 2500 psia pressure.

Reference 2 addresses the Model 93 RCP main closure bolts, which are relatively unaffected by system thermal transients since they are above the thermal barrier cooling coils and, hence, operate at a reduced temperature. Reference 3 addresses the main closure flange, which is also at less than the system temperature and above the thermal barrier. It is concluded that the system transients have a negligible effect on the closure flange.

Reference 4 presents the analysis of the Model 93 RCP. Reference 4 has a normal operating case analysis and also bounds the anticipated transient without trip (ATWT). This analysis covers the casing, closure flange and bolts, the thermal barrier heat exchanger, and the seal housing and bolts. The ASME Code limits were met for the ATWT as an Emergency (Level C) transient, where only primary stresses are considered. Reference 5 shows the Model 93 RCP casing foot meets paragraph 4.3.7 of the E-Spec.

Reference 6 covers the seismic analysis of the Model 93 RCP for lateral and vertical static g-loads, which shows no yielding in critical components.

NSSS Design Transient Evaluation The new NSSS design transient data for the uprate RCS parameters were compared to the original E-Spec values. Cycle counts in general are the same except for a few cases that decrease. The design transients also introduce some new transients to be considered. In comparing transient changes, temperature increases (+AT) over normal operating temperature and pressure increases (+AP) over normal operating pressure are of most interest. Decreases represent lower stresses than heatup-cooldown and, hence, do not affect the stress intensity range. The transient magnitudes can be shown to have no impact on fatigue by showing the AT, AP values do not qualify as a significant fluctuation per N-415.1 of the fatigue waiver requirements in the ASME Code. Thus, they would not be included in a fatigue waiver analysis or be considered to have any fatigue usage associated with the transient. Those cases of interest are as follows.

07/30/02 5-2

1. Large Step-Load Decrease: The revised transient has a slightly larger AT (5°F more) and AP (8.7 psi more) than shown in the original E-Spec. Both increases are less than the significant fluctuation values calculated in Items 2 and 3 below. The maximum is less than the design pressure, hydro-pressure, and ATWT pressure cases previously analyzed in References 2, 3 and 4.

2. Feedwater Cycling: This is a new transient with no pressure increase and only a 5°F temperature increase. There are 25,000 cycles of AT = 32'F, but this is not significant fluctuation per the ASME Code. From N-415.1 (d), a significant temperature fluctuation is given by:

Significant temperature fluctuation, AT Sa- 26000 -51.6 0F 2Ea 2(25.8)(9.76) where Sa allowable amplitude of alternating stress intensity from the design fatigue curve given in the ASME Code for the material under consideration. For 304SS at 550'F, Sa at 106 cycles = 26,000 psi.

E modulus of elasticity for the material under consideration from Table N-427 of the ASME Code. For 304SS at 550'F, E = 25.8 x 106 psi.

c = the instantaneous coefficient of thermal expansion of the material under consideration at the mean temperature under consideration. For 304SS at 310'F, a = 9.76 x 10-6 in./in./°F.

Thus, the 32°F fluctuation is not a significant fluctuation to include in a fatigue waiver analysis.

3. Boron Concentration Equalization: This is a new transient with a high number of cycles but of low magnitude of pressure (33 psi). A significant pressure fluctuation per N-415.1(b) of the ASME Code is given by:

1 x 5I Significant pressure fluctuation, AP = Design Pressure x -

3 Sm where:

S = Sa at 106 cycles Sa allowable amplitude of alternating stress intensity from the design fatigue curve given in the ASME Code for the material under consideration. For 304SS at 550'F, Sa at 106 cycles = 26 ksi.

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Sm allowable design stress intensity for the material under consideration from Table N-421, N-422, or N-423 of the ASME Code. Sm= 16.9 ksi for most 304SS at 550'F, and somewhat higher for the 316 SS casing. It is noted that the value of Sm for 304SS at 550'F was smaller than 16.9 ksi prior to the Winter 1969 Addenda, but the newer value is used as it is more conservative in this equation than the earlier value.

Significant pressure fluctuaton, AP = 2500 - 16 = 1282 psi Thus, this transient does not qualify as a significant pressure fluctuation to be included in a fatigue waiver and, thus, has no fatigue usage.

4. Loss of Power: This is a new transient which also does not have significant fluctuations, i.e., AT = 20'F < 51.6°F and AP = 250 psi < 1282 psi. Thus, this transient does not have any fatigue significance.

5. Loss of Flow: This transient has a SG pressure increase of 59 psi whereas previously none was defined while the AT is less. Again, this is not a significant pressure fluctuation and, thus, is of no concern for fatigue.

6. Inadvertent Actuation of Auxiliary Spray: This is a new upset case transient that has a pressure decrease of 1200 psi. This is less than the significant AP value previously calculated and, hence, has no significance in fatigue.

7. Steam Line Break: This is a new faulted case transient with only temperature and pressure decreases. Hence, the primary stress intensities decrease. Faulted conditions are not included in a fatigue evaluation.

8. Turbine Roll Test: This is a new test condition with a pressure and temperature decrease. The ASME Code had no testing stress analysis criteria then. Since the pressure only decreases, there would be no primary stress concerns.

The changes in the transients due to the revised RCS conditions were evaluated as described above and found to be acceptable for the Model 93 RCPs. No increase in fatigue usage is anticipated due to the uprating transients.

5.4.2 Reactor Coolant Pump Motor The RCP motor performance was evaluated based on the performance characteristics of the original impellers for the worst-case condition, which is the uprated NSSS power level, reduced thermal design flow (TDF), and a bounding 25-percent steam generator tube plugging. Loads on the RCP motors based 07/30/02 5-4

on a steam generator outlet temperature of 525.7°F and a best-estimate flow of 87,200 gpm were calculated. The results show a hot loop motor load of 5473 HP and a cold loop motor load of 7173 HP.

The PBNP RCP motors (serial numbers 1S-75P343, 2S-75P343, IS-75P344, and 2S-75P344) have a nameplate rating of 6000 HP. Since the loads are less than the nameplate rating of the motors (the cold loop rating is taken to be 125 percent of the hot-loop rating if not shown separately), no analysis is necessary for operation at these conditions. Per the original design specification, the motors are acceptable for any load up to 6000 HP hot loop and 7500 HP cold loop.

5.5 Conclusions Based on the evaluation, the design of the Model 93 RCP meets the applicable ASME Code requirements for structural integrity at the revised RCS conditions associated with the uprated core power of 1650 MWt (1656 MWt NSSS power) for up to 60-year life. The motors are also acceptable for the loads calculated for the uprated RCS conditions.

5.6 References

1. Westinghouse Equipment Specification 676433, Rev. 1, "PBNP Station No. I and No. 2 Reactor Coolant Controlled Leakage Pump," R. C. Moren, October 9, 1967.

2. Engineering Memorandum 3714, Revision 2, "Main Flange Bolted Joint Stress Analysis,"

Westinghouse Electro-Mechanical Division, B. Cuerden, May 2, 1976.

3. Engineering Memorandum 3747, "Main Flange Shell Stresses," Westinghouse Electro-Mechanical Division, F. Van Alen, July 7, 1966.

4. Engineering Memorandum 4892, "Structural Analysis of 93 RCP Casing and Closure for ATWT Event," Westinghouse Electro-Mechanical Division, B. Cuerden, October 15, 1976.

5. Engineering Memorandum 3648, "Casing Foot Stresses," Westinghouse Electro-Mechanical Division, A.P. Villasor, Jr., March 11, 1965.

6. Engineering Memorandum 3992, "Static Seismic Load Stress Analysis," Westinghouse Electro-Mechanical Division, by F.K. Van Alen, July 30, 1968.

7. ASME Boiler and Pressure Vessel Code, Section 111-NB and Appendices, 1968 Issue and Summer 1968.

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6.0 Steam Generators 6.1 Introduction The PBNP has two different Westinghouse steam generator models installed. Unit I has Model 44F steam generators and Unit 2 has Model A47 steam generators. Each of these models was analyzed for the uprated power conditions with respect to structural acceptability and thermal-hydraulic (T/H) performance.

Evaluations were performed to address the 8.7 percent power uprate to a core power of 1650 MWt (1656 MWt NSSS power). The evaluations considered plant operation up to 60 years with renewed operating licenses.

6.2 Key Input Parameters The evaluation of the uprated power used the NSSS design parameters and revised NSSS design transients.

6.3 Acceptance Criteria Structural The critical steam generator components that were evaluated for structural adequacy are:

Primary side: Primary chamber, tubesheet, primary nozzles, primary manway, divider plate, and tube-to-tubesheet weld. The primary side of the replacement steam generators was evaluated as a whole through a review of the uprating transients that affect the primary side of the steam generator, i.e., RCS transients.

Secondary side: Upper shell, transition cone, lower shell, junction of tubesheet and stub barrel, main and auxiliary feedwater and spray nozzles, secondary manway opening and bolts, inspection ports, and minor shell taps.

These components were evaluated for the effects of the uprate on the steady-state and transient conditions for the normal and upset loads in the design specifications, References 1 (Model 44F) and Reference 2 (Model A47). The test, emergency, and faulted loading conditions are unaffected by the uprate. The structural acceptance criteria for both steam generator models are given in the 1965 Edition through Summer 1966 Addenda of the ASME B&PV, Section If, Reference 3. Details of the actual acceptance criteria employed in the structural evaluation of both the 44F and A47 are given in Section 4 of Volume 1 of Reference 4.

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Thermal-Hydraulic The acceptability of the thermal-hydraulic performance of the steam generators is assessed by evaluating the impact of the uprated operating conditions on several key secondary-side operating characteristics.

To determine the acceptability it is necessary to evaluate the changes in these key parameters, then assess how the combined changes affect the overall thermal hydraulic performance when operating at the 1656 MWt power level. With the exception of moisture carryover (MCO), the individual performance parameters are not compared to previously established limiting values, rather, acceptability is based on the magnitude of change in the parameter. For MCO, the established limit is a maximum of 0.25 percent.

This limit is based on the operating characteristics of the downstream components (e.g., turbines).

Thermal-Hydraulic Operating Characteristics Considered Circulation Ratio The circulation ratio (CR) is a measure of liquid flow in the bundle in relation to the steam flow. It is primarily a function of power. The bundle liquid flow must remain high enough to prevent the accumulation of contaminants in the tube bundle and on the tubesheet.

Damping Factor The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit.

Steam Generator Mass The steam generator mass relates to the energy release in case of an accidental event. A small increase in steam generator mass should be acceptable, unless there is a definite specified limit stated in the safety analysis.

Peak Heat Flux The peak heat flux should be below the predicted value at departure from nucleate boiling (DNB) transition.

Steam Gen. Pressure Drop The increase in the total secondary-side AP for the steam generator should be small compared to the total feed system AP. The pressure drop is small compared with the secondary side bundle pressure. A small change in pressure drop would not affect the T/H performance of the steam generator.

Moisture Carryover The projected MCO should not exceed the design limit of 0.25 percent.

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6.4 Description of Analysis/Evaluation and Results 6.4.1 Structural Evaluation - Model 44F The Westinghouse Model 44F steam generators were installed as replacements for the original Model 44 steam generators at PBNP Unit 1.

6.4.1.1 Structural Evaluation Methodology - Model 44F The Model 44F steam generators, currently installed in Unit 1, have been evaluated using the NSSS design parameters for operation at power uprate. The maximum pressure differential (AP) between the primary side and the secondary side must not exceed the design pressure differential of 1550 psi under normal operating conditions, per the equipment specification. Therefore, when the primary-side pressure is 2250 psia, the steam pressure on the secondary side must not fall below 700 psia, not including the pressure variations due to the design transients. When the design transient pressure variations are considered, it will be necessary to limit secondary-side steam pressure to a minimum of 745.7 psia to achieve a maximum AP of 1550 psi. The secondary-side pressure limitation impacts only the low Tavg temperature operation. In actual operation, the secondary-side minimum pressure will be limited to ensure the 1550 psi AP is not violated.

Uprating conditions were compared to the originally analyzed conditions with no uprating. Scaling factors were developed for the following based on this comparison: primary and secondary temperature, secondary-side pressure and primary-to-secondary pressure differential. The existing pressure induced steady-state stresses will be adjusted by the appropriate scaling factor to obtain the stresses for the uprated conditions.

For the Model 44F steam generators, the design transients and applicable cycles are postulated for a total of 60 years of operation. This is an additional 20 years of operation beyond the currently licensed 40 years. The load cycles for the currently licensed period are given in the original design specification, Reference 1. The "50-percent Load Decrease" transient in the original design specification has been renamed the "Large Step-Load Decrease." An evaluation of several significant transients was performed for the uprated conditions. An evaluation of the temperature differences, range, and gradients was made for the following major bounding transients: 10-percent load increase/decrease, Large step-load decrease (same as 50-percent step-load decrease), and reactor trip. A comparison with the transients used in the original stress report, Reference 5, indicates that primary-side temperature variations are somewhat greater for the uprating cases. However, the thermal gradients across the thickness of the steam generator components do not change significantly. The secondary-side transients basically remain unchanged.

The primary-to-secondary pressure differential changes were calculated at the uprated conditions for these same three transients and compared to the transients used in the original stress report. Scaling factors were developed based on this comparison. The effect of these changes is reflected mainly in the stress range calculations. In the stress range calculations, the stress range determined in the original 07/30/02 6-3

stress report is adjusted by the appropriate scaling factor to determine the limiting stress range for the uprating.

The factors that have been evaluated are conservatively applied to the maximum primary plus secondary stress range calculations to ensure that the ASME Code criteria (Reference 3) are met. For the applicable cases, the appropriate factors have been applied to the alternating stresses in order to update the fatigue evaluation, to ensure that the cumulative fatigue usage stays below unity per the requirements of the ASME Code. It has been concluded that all the components are adequate for 60 years of operation with the exception of inspection port bolts, as discussed in the next section.

6.4.1.2 Results of the Structural Evaluation - Model 44F Critical steam generator components, evaluated structurally, are the channel head, the tubesheet, the tubesheet junctions, the tube-to-tubesheet weld, the tubes, the secondary shell, minor shell penetrations, and the feedwater nozzle. Although the divider plate is not an external pressure boundary component, it is also evaluated. Somewhat higher pressure drops act across the divider plate at increased tube plugging levels. In addition, the structural evaluation of the relocated narrow range level taps for the uprated conditions and 60 years of operation is included. The results of the evaluation are valid for the uprate to 1656 MWt NSSS power. The fatigue usage summary for the Model 44F steam generator is presented in Table 6-1.

Channel Head, Tubesheet, and Tubesheet-to-Shell Junctions - Model 44F The channel head, the tubesheet, and the tubesheet's junctions with the secondary shell and the channel head are included in the finite element model developed and used in the primary components interaction analysis included in Reference 5. These were modified for uprated conditions. Design, LOCA, and test condition stresses remain unchanged. Stress results for primary plus secondary stress intensity range and fatigue usage for normal and upset conditions were evaluated. All ASME Code stress limits are satisfied.

The calculated fatigue usage values for the normal power condition from Reference 5 are very small.

Therefore, this conclusion, with respect to fatigue, remains valid for the uprated conditions. The calculated usage factors are expected to remain small and within the ASME Code allowable limit of one, even if the uprate and the augmented load cycles for the 60 years of operation are considered.

Tube-to-Tubesheet Weld - Model 44F Summary stress results for the tube-to-tubesheet weld are given in Reference 5 for current power rating.

These results are modified to reflect the changes that occur due to the uprated conditions. The stress results for primary plus secondary stress intensity range for normal and upset conditions were evaluated.

The ASME Code stress limit is satisfied. Design, LOCA, and test condition stresses remain unchanged.

The calculated fatigue usage at the most critical location for the uprated conditions and 60 years of operation is less than 1.0. This fatigue usage value was calculated using the finite element results from 07/30/02 6-4 07/30102 6-4

Reference 5 plus a conservative additional fatigue strength reduction of 4.0, and is within the maximum allowable limit of 1.0.

Tubes The summary stress results for the tubes are given in Reference 5 for current power rating. The transients for the uprated conditions were reviewed against the transients used for the normal power condition. The effect of these new transients on the tubes remains unchanged. Therefore, the existing results remain valid. All ASME Code stress limits are satisfied. The fatigue analysis calculations were revised considering the augmented load cycles for 60 years of operation. Both the primary-plus secondary stress and usage factor are within the maximum allowable limits.

Secondary Shell - Model 44F Summary stress results for the secondary shell transition cone are given in Reference 5 for current power rating. These results remain bounding for the uprated conditions since a reduction in secondary pressure will reduce the stresses in the shell. All ASME Code stress limits are satisfied. For fatigue, the overall governing location for the secondary shell has already been considered in the evaluation for the channel head, the tubesheet and the tubesheet to shell junctions. The structural evaluation of the relocated PBNP Unit 1 level taps in the secondary shell is discussed below.

Minor Shell Penetrations - Model 44F Summary stress results for minor shell penetrations are given in Reference 5 for current power rating conditions. These results are modified for the uprated conditions. All ASME Code stress limits are satisfied. For fatigue, the inspection port bolts are the most critical. To keep the calculated usage factor less than one, the inspection port bolts should be changed every 12 years for uprated conditions instead of the current replacement interval of 10 years. Other penetrations are exempt from fatigue calculations.

Feedwater Nozzle - Model 44F The original analysis in Reference 5 was performed with a feedwater temperature of 436.5 0 F (Reference 1). For the uprate to an NSSS power of 1656 MWt, the feedwater temperature will increase.

The increase in temperature reduces the temperature gradiant across the nozzle and will result in reduced thermal stresses.

The secondary-side pressure reduction for the uprate will improve the pressure (primary) stress conditions in the nozzle. The stress ratio results for the feedwater nozzle, obtained for the current power rating, remain applicable for the uprated conditions. All ASME Code stress limits remain satisfied.

Augmented load cycles for the postulated 60 years of operation will not significantly change the fatigue usage values, which remain within the maximum allowable limit. Feedwater flow stratification is a flow phenomenon whereby stagnant temperature gradients occur at different cross sections in a pipe. The 07/30/02 6-5

difference in metal temperature may result in fatigue issues in the pipe wall. This issue is primarily controlled by the geometry of the piping. The conditions relevant to feedwater flow stratification are not affected by the proposed uprating.

Divider Plate - Model 44F The divider plate is welded to the tubesheet and channelhead over its perimeter to maintain separation of hot-leg and cold-leg fluids. Since the divider plate is not an external pressure boundary, it is a Class 2 component, Reference 1. Even though not strictly required, more stringent ASME Code Class I acceptance criteria for pressure boundary components are used as a guide for the analysis of the divider plate to show its structural adequacy. For the tube plugging level of 25 percent (bounding) at the uprated power condition, a somewhat higher differential pressure will act across the divider plate. The increase in pressure will produce an additional bending stress at the divider plate to tubesheet junction.

The effect of this additional stress however, is not significant. The resulting primary stress intensities are well within the primary allowable stresses and the 3Sm limit of 70 ksi at 600'F for primary plus secondary stress intensity range continues to be met. In the fatigue usage factor calculations, a stress concentration factor of 2.0 is used for obtaining peak stresses at the critical junction and the postulated load cycles are augmented for 60 years of operation. The resulting maximum calculated usage factor is less than 1.0 and is less than the ASME Code fatigue usage limit.

Upper Shell Remnant - Model 44F The upper shell (along with its manway) and the steam outlet nozzle are remnant components from the original 44 Series steam generator. The remnant components were evaluated for continued use in Model 44F replacement steam generators in Reference 5. The power uprate results in reduced secondary (steam) pressures and temperatures. Therefore, the specified loads, considered in Reference 5, bound the structural evaluation. The calculated fatigue usage factor for 40 years is less than 1. Since the maximum usage in the remnant based on 40 years is very low, extension to 60 years and ASME Code compliance within the usage limit of one is acceptable.

Level Tap Relocation - Model 44F The narrow-range lower level taps were relocated from the upper cylinder about 63 inches toward the tubesheet, which places them in the transition cone, about 22 inches above the junction with the lower cylinder. A structural evaluation of the relocated lower level taps was performed. The analysis considered a conservative set of secondary side loading conditions given in the original 44F design specification, Reference 1, which bound the relatively lower steam pressures associated with the uprate conditions. The tap holes are spaced sufficiently apart to meet the ASME Code requirements for un-reinforced openings. The primary stress limits are not affected by the relocated new tap holes, since the local stresses in the conical shell due to the holes are secondary. Therefore, the existing positive primary stress margins in the conical shell for design, faulted and test loads remain valid and bounding.

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Three-dimensional finite element analysis of the local shell and tap penetrations was performed to calculate the maximum stresses due to the various specified pressure and thermal transient loads. The resulting overall bounding maximum secondary stress range at the new tap holes is less than an allowable of 90 ksi. The maximum cumulative fatigue usage factor calculated for 40 years (as specified in Reference 1) is less than 1.0. This factor was calculated using several very conservative assumptions, which if removed, would result in a fatigue usage factor of less than one for 60 years.

Flow-Induced Tube Vibration and U-Bend Fatigue - Model 44F In addition to the assessment to demonstrate satisfaction of the Section III ASME Code structural requirements, an evaluation of the potential for high cycle fatigue of unsupported U-bend tubes was performed. One of the prerequisites for high cycle U-bend fatigue is the formation of a dented support condition at the upper plate. This support condition is a result of a build up of corrosion products associated with drilled holes in carbon steel tube support plates (TSPs). Since the broached stainless steel support plate is designed to inhibit the introduction of corrosion products, the support condition necessary for the development of high cycle fatigue cannot occur. As a result, high cycle fatigue associated with unsupported inner row tubes also cannot occur in this model steam generator.

With respect to tube wear, the baseline analysis for the Point Beach Unit 1 Model 44F steam generators indicates that wear as a result of tube vibration is very small over the projected design life of the unit.

The rate of tube wear resulting from the proposed power uprate (to 1656 MWt) has been determined to increase after the uprate, but not significantly. The maximum wear for a 40-year operating life of the replacement steam generators has been projected to increase from -3 mils (original pre-uprate condition) to less than -5 mils (post-uprate). This conservatively assumes that the uprate is over the entire 40 year life. Extending this to 60 years, the expected wear would be 1.5 times that for 40 years or less than 8 mils (nominal). With a typical plugging limit of 40-percent through-wall, (20 mils for a wall thickness of 0.050 in.), sufficient margin exists to account for the wear anticipated due to vibration subsequent to the implementation of the uprate for the 60-year period of plant operation.

From the above, it has been determined that the limiting aspects of flow induced vibration associated with steam generator tubing, (fatigue and tube wear) will not be affected by the proposed uprate to 1656 MWt.

Summary Results - Model 44F Results of analyses described above for the Unit 1 Model 44F steam generator components show that the structural integrity of the steam generator components would be maintained for the 1656 MWt (NSSS power) uprated operations with a maximum plugging level of 10 percent in the steam generator tubes.

The fatigue calculations were performed for augmented load cycles postulated for 60 years of operations.

The uprating conditions with increased load cycles were accommodated without exceeding the ASME Code limits. It has been concluded that all components are adequate for 60 years of operation with the 6-7 07/30/02 07/30/02 6-7

exception of the inspection port bolts, which need to be replaced every 12 years. The fatigue usage summary for the Model 44F steam generator is presented in Table 6-1.

6.4.2 Structural Evaluation -Model A47 Westinghouse Model A47 steam generators were installed as replacements for the original Model 44 steam generators at PBNP Unit 2. The Model A47 design specification is documented in Reference 2.

The Model A47 steam generators were originally qualified for 1650 MWt NSSS power, with up to 25-percent SGTP. The external and support interface dimensions of the Model A47 steam generators are essentially identical to those of the original Model 44 steam generators with only minor variations.

These variations are attributed to enhanced materials and manufacturing tolerances with a minor increase in the secondary shell barrels inside diameter, associated with decreased shell wall thickness and machined inside/outside cylinders. The outside diameter of the channel head, tubesheet, lower, and upper shell barrels remain within the maximum permitted by the original Model 44 steam generator design.

Like the Model 44 and Model 44F steam generators, the Model A47 is a feedring type steam generator with the main feedwater nozzle located in the upper shell. The feedwater inlet nozzle is truncated to pass through the existing reactor building equipment hatch and weld reassembled inside containment to interface with existing plant equipment. The lower narrow range water level taps have been relocated to the shell cone to enhance high- and low-level trip response times relative to those of the original Model 44 steam generator.

The Model A47 steam generator design embodies the key characteristics of the Westinghouse Model F design, namely, the use of thermally treated Alloy 690 tube material, Type 405 stainless steel tube support material, broached flat contact tube supports, hydraulically expanded tubesheet joints, and "minimum gap" U-bend constructions.

6.4.2.1 Structural Evaluation Methodology - Model A47 The design basis for the Model A47 steam generators incorporates the NSSS design parameters for the 1656 MWt NSSS power uprate conditions. The design specification (Reference 2), for the PBNP Unit 2 Model A47 steam generators, satisfies the structural requirements of Section III of the ASME Code, Reference 3, for conditions that include the uprating to 1656 MWt. The final structural analysis is documented in the PBNP Unit 2 Model A47 Stress Report, Reference 4. In general, direct finite element simulation of the various Model A47 steam generator components was employed to calculate pressure and thermal transient stresses, as discussed in Reference 4. Results of evaluation show that the structural integrity of the steam generator components would be maintained for the 1656 MWt uprated operation.

All components are adequate for 60 years of operation 6-8 07/30102 07/30/02 6-8

6.4.2.2 Results of the Structural Evaluation - Model A47 Evaluation results for the major A47 components are summarized below. The fatigue usage factors reported are based on the numbers of occurrences specified in Reference 2 for 40 years. With few exceptions, based on 20 years of actual service at PBNP Units 1 and 2, the number of anticipated occurrences for even 60 years of plant life was less than those originally specified for 40 years. The LOCA, design basis earthquake (DBE), and test conditions are independent of the uprate. With respect to nominal and uprated power conditions, the results in Reference 4 consider "Normal and Abnormal" loads, which are usually referred to as "normal (level A) plus upset (level B)" load conditions. The other "abnormal" condition considered in Reference 4 is the "loss of shell side pressure" which is treated as an emergency (level C) condition with respected to the ASME Code stress limits in the design specification, Reference 2.

All Model A47 steam generator components meet all structural criteria of the ASME Code, Reference 3, for the load conditions specified in Reference 2. The fatigue usage summary for the Model A47steam generator is presented in Table 6-2.

Channel Head, Tubesheet, and Tubesheet-to-Shell Junctions - Model A47 The channel head, the tubesheet, and the tubesheet's junctions with the secondary shell and the channel head are included in the finite element model developed and used in the primary components interaction analysis included in Reference 4. Stress intensity range and fatigue usage for normal and upset conditions were evaluated. All ASME Code stress limits are satisfied. The calculated fatigue usage factor at the most critical location for the uprated conditions remains within the ASME Code allowable limit of one.

Tube-to-Tubesheet Weld - Model A47 Summary stress results for the tube-to-tubesheet weld are given in Reference 4. Stress intensity range and fatigue usage for normal and upset conditions were evaluated. All ASME Code stress limits are satisfied. The calculated fatigue usage factor at the most critical location for the uprated conditions remains within the ASME Code allowable limit of one.

Tubes - Model A47 Summary stress results for the tubes are given in Reference 4. The results show that all ASME Code stress limits are satisfied including external pressure limits and thermal stress ratcheting limits. The calculated fatigue usage factor for the uprated conditions remains within the ASME Code allowable limit of one.

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Secondary Shell - Model A47 Summary stress results for the secondary shell are given in Reference 4. These results are bounding for the uprating conditions, which occur with a reduction in secondary pressure and result in smaller primary stresses in the shell. All ASME Code stress limits are satisfied. The calculated fatigue usage factor at the most critical location for the uprated conditions remains within the ASME Code allowable limit of one.

Minor Shell Penetrations and Upper Internals - Model A47 Summary stress results for minor shell penetrations and the upper internals are given in Reference 4. All stress limits are satisfied for all minor shell penetration and upper internals components. The calculated fatigue usage factor for the uprated conditions remains within the ASME Code allowable limit of one.

Feedwater Nozzle and Thermal Sleeve - Model A47 Summary stress results for the Model A47 feedwater nozzle and thermal sleeve are given in Reference 4.

All ASME Code stress limits are satisfied. The calculated fatigue usage factor at the most critical location for the uprated conditions remains within the ASME Code allowable limit of one.

Divider Plate - Model A47 The Model A47 divider plate is welded to the tubesheet and channel head over its perimeter to maintain separation of hot-leg and cold-leg fluids. Since the divider plate is not an external pressure boundary, it is a Class 2 component, Reference 2. Even though not strictly required, more stringent ASME Code Class 1 acceptance criteria for pressure boundary components are used in Reference 4. All pressure stress limits are satisfied. However, the 3 Sm elastic limit was exceeded and a plastic analysis was performed, as permitted by Section N-417.5 of the ASME Code, Reference 3. The resulting maximum calculated fatigue usage factor is less than the ASME Code limit of one.

Upper Shell, Head, and Steam Nozzle - Model A47 All ASME Code stress limits are satisfied. The calculated fatigue usage factor at the most critical location for the uprated conditions remains within the ASME Code allowable limit of one.

Flow-Induced Tube Vibration and U-Bend Fatigue - Model A47 In addition to the assessment to demonstrate satisfaction of the Section III ASME Code structural requirements, an evaluation of the potential for high cycle fatigue of unsupported U-bend tubes was performed. One of the prerequisites for high cycle U-bend fatigue is the formation of a dented support condition at the upper plate. This support condition is a result of a build up of corrosion products associated with drilled holes in carbon steel TSPs. Since the broached stainless steel support plate is 07/30/02 6-10

designed to inhibit the introduction of corrosion products, the support condition necessary for the development of high cycle fatigue cannot occur. As a result, high cycle fatigue associated with unsupported inner row tubes also cannot occur in this model steam generator.

With respect to tube wear, the baseline analysis for the Point Beach Unit 2 Model A47 steam generators indicates that wear as a result of tube vibration is very small over the projected design life of the unit.

The rate of tube wear resulting from the proposed power uprate (to 1656 MWt) has been determined to increase after the uprate, but not significantly. The maximum wear over a 40-year operating life of the replacement steam generators has been projected to increase from -2.6 mils (original pre-uprate condition) to less than -5 mils (post-uprate). With a typical plugging limit of 40 percent through-wall, (20 mils for a wall thickness of 0.050 in.), sufficient margin exists to account for the wear anticipated due to vibration subsequent to the implementation of the uprate.

From the above, it has been determined that the limiting aspects of flow induced vibration associated with steam generator tubing, (fatigue and tube wear) will not be affected by the proposed uprate to 1656 MWt NSSS power.

Summary Results - Model A47 Results of analyses described above for the Unit 2 Model A47 steam generator components show that the structural integrity of the steam generator components would be maintained for the 1656 MWt (NSSS power) uprated operations with a maximum plugging level of 10 percent in the steam generator tubes.

The fatigue calculations were performed for augmented load cycles postulated for 60 years of operations.

The uprating conditions with increased load cycles were accommodated without exceeding the ASME Code limits. It has been concluded that all components are adequate for 60 years operation. The fatigue usage summary for the Model A47 steam generator is presented in Table 6-2.

6.4.3 Thermal-Hydraulic Evaluation - Model 44F And Model A47 The thermal- hydraulic performance of the PBNP Model 44F and Model A47steam generators have been evaluated at the uprated conditions up to 1656 MWt NSSS power.

6.4.3.1 Analyses and Evaluations Performed The thermal-hydraulic evaluations assessed steam generator operating characteristics, including moisture carryover. Attention was focused on secondary-side parameters. Parameter values calculated for the uprated conditions are compared to the values at the non-uprated design conditions. Where appropriate, the parameter values are compared to other existing field experience. A discussion of moisture separator performance is included.

07130102 6-11 07130/02 6-11

6.4.3.2 Results - Model 44F And Model A47 Thermal-Hydraulic Operating Characteristics Several secondary-side operating characteristics are used to assess the acceptability of steam generator operation at uprated conditions.

Circulation Ratio The CR is a measure of liquid flow in the bundle in relation to the steam flow. It is primarily a function of power. At the uprated NSSS power, the circulation ratio decreases by a small amount. Since the steam flow increases with power, the bundle liquid flow also decreases. However, the bundle liquid flow is expected to be large enough to minimize the accumulation of contaminants on the tubesheet and in the bundle. The uprating and plugging level, therefore, have no major effect on this function.

Damping Factor The hydrodynamic stability of a steam generator is characterized by the damping factor. A negative value of this parameter indicates a stable unit; that is, small perturbations of steam pressure or circulation ratio will die out rather than grow in amplitude. At uprated conditions the damping factors become slightly more negative. Therefore at uprated conditions the generators will continue to be hydrodynamically stable.

Steam Generator Mass The reduced steam pressure, which results from plugging, primary temperature reduction, and to a lesser degree from the uprating, brings about an increased void fraction in the tube bundle. This results in a small decrease in steam generator mass with NSSS power uprating. The small reduction will have no effect on processes related to voids in the tube bundle. Also since the secondary side mass is related to energy release during a transient event, such as a steam line break, the reduction in mass is less limiting.

As such there is no impact on steam generator performance.

Peak Heat Flux The value of peak heat flux will increase with power and tube plugging. For uprating, increased total heat load is passed through the same bundle heat transfer area, increasing the heat flux. For increased plugging, the same heat load is passed through a smaller heat transfer area, also increasing the heat flux.

Sufficient margin remains to ensure that departure from nucleate boiling (DNB) does not occur at uprated conditions with up to 10-percent SGTP.

07/30/02 6-12

Steam Pressure It will be necessary for the plant to maintain steam pressure above 745.7 psia for all operating conditions.

Predictions indicate that this value is approached only at the low end of the Tavg range with the maximum allowable tube plugging.

Steam Generator Pressure Drop For the power uprate, the total secondary-side pressure drop increases by 6.2 psi, which is small in comparison to total pressure drop in the feedwater system. Thus, the increase in secondary-side pressure drop due to the uprate should have no significant effect on the feedwater system operation.

Moisture Carryover Field tests for moisture carryover of Model 44F steam generators have been performed at PBNP Unit I and at other plants with the same separator package. These results show that the separator packages are highly effective. Measured carryover is near 0.01 percent at full power. Using available data, it is possible to evaluate the expected change in moisture carryover at low steam pressures and uprated conditions. Component test results at uprated power conditions have also been considered. The evaluation demonstrates that continued high performance for the separator packages at PBNP Unit 1 is expected for much of the range of conditions being evaluated but may become marginal in terms of a 0.25-percent limit for the uprated power, especially at low steam pressure. The Unit 2 Model A47 steam generators have the same separator and dryer packages as are installed in the Model 44F steam generators. The moisture separator package for each steam generator model is expected to deliver a moisture of less than 0.25 percent at the 1656 MWt NSSS power level if a steam pressure above 745.7 psia is maintained.

6.5 Conclusions Results of analyses described above for the Unit I Model 44F and the Unit 2 Model A47 steam generator components show that the structural integrity of the steam generator components would be maintained for the indicated uprated operations at NSSS power levels up to 1656 MWt, with a maximum plugging level of 10 percent in the steam generator tubes. The fatigue calculations were performed for augmented load cycles postulated for 60 years of operations. Increased load cycles were accommodated without exceeding the ASME Code limits. The lone exception is the Unit 1 Model 44F steam generator inspection port bolts, which need to be replaced every 12 years.

The thermal-hydraulic operating characteristics of the PBNP Unit 1 Model 44F, and Unit 2 Model A47 generators are within acceptable ranges for anticipated uprated conditions to 1656 MWt. For both the Unit 1 Model 44F and the Unit 2 Model A47 generators it is predicted that the MCO would approach the 0.25 percent limit when operating in the uprated conditions at low steam pressures. Maintaining steam pressure above 745.7 psia will provide margin for the MCO to remain below the 0.25 percent limit.

07/30/02 6-13

Calculations show that theoretically the steam pressure could fall below the plant minimum limit of 745.7 psia when the Model 44F and A47 steam generators are operated with a low primary side T,,g and 10-percent SGTP at the uprated power condition.

6.6 References

1. Wisconsin Electric Power Company Point Beach Nuclear Plant Unit 1 (Model 44F), Design Specification 955381, Revision 2, September 16, 1983.

2. Wisconsin Electric Power Company Point Beach Nuclear Plant Unit 2 (Model A47), Design Specification 412A72, Revision 2, November 22, 1996.

3. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Vessels, 1965 Edition Through Summer 1966 Addenda, American Society of Mechanical Engineers, New York, N.Y.

4. WNEP-9513, "Model A47 Steam Generator Stress Report, Summary Stress Report," Wisconsin Electric Power Company, Point Beach Unit 2, Revision 1, December 1996.

5. "Model 44F Replacement Steam Generator Stress Report for Wisconsin Power Company - Point Beach Unit I ," Stress Report No. WNEP-8393, Westinghouse Electric Corporation, October, 1983.

6-14 07/30/02 6-14

Table 6-1 Summary of Fatigue Usage PBNP Unit 1 - Model 44F Steam Generator Components Components Fatigue Usage Primary Side Components [ ]bcO)

(Max. of all components - channel head, tubesheet, tubesheet-to-shell junction and secondary shell)

Tube-to-Tubesheet Weld [ ]b,c Tubes I]b'c Secondary Shell Transition Cone [ ]bc()

Minor Shell Penetrations - Excluding Bolts [ ]bc Minor Shell Penetrations - Inspection Port Bolts [ ]bc(2)

Main Feedwater Nozzle [ bx(3)

Divider Plate [ ib,c Upper Shell Remnant - Manway Pad Ibx Level Tap Relocation [ ]bC Notes:

1. Fatigue usage factor listed represents 40 years of operation. The value listed in the table is significantly below the allowable of 1.0 that extending the plant life based on the cycles postulated for 60 years of operation will not significantly affect the fatigue usage and component qualification.

2. To keep the usage factor less than one, the inspection port bolts must be replaced every 12 years of service at the uprated conditions.

3. Fatigue usage factor was evaluated to be less than 1.0. The previous calculated usage factor the main feedwater nozzle was [ ]b,,.

Bracketed [ ] b,, information designates data that is Westinghouse Proprietary.

07/30/02 6-15 07/30/02 6-15

Table 6-2 Summary of Fatigue Usage PBNP Unit 2 - Model A47 Steam Generator Components Components Fatigue Usage Primary Side Components [ b,c (Max. of all components - channel head, tubesheet and tubesheet-to shell junction)

Tube-to-Tubesheet Weld [ ]b,c Tubes [ ]btc Secondary Shell Transition Cone [ ]bc Minor Shell Penetrations And Upper Internals [ ]b,c Main Feedwater Nozzle And Thermal Sleeve [ bc Divider Plate [ ]b.c Upper Shell, Head And Steam Nozzle I ]b.C Bracketed [ ] b,c infonnation designates data that is Westinghouse Proprietary.

6-16 07/30/02 6-16

7.0 Pressurizer Evaluations were performed to assess the impact on the pressurizer components of a power uprate for PBNP Units 1 and 2 to a core power of 1650 MWt for up to 60 years with renewed operating licenses.

7.1 Introduction The functions of the pressurizer are to absorb any expansion or contraction of the primary reactor coolant due to changes in temperature and pressure and to keep the RCS at the desired pressure. The first function is accomplished by keeping the pressurizer approximately half full of water and half full of steam at normal operating conditions, connecting the pressurizer to the RCS at the hot leg of one of the reactor coolant loops and allowing inflow to or outflow from the pressurizer as required. The second function is accomplished by keeping the temperature in the pressurizer at the water saturation temperature (Tsat) corresponding to the desired pressure. The temperature of the water and steam in the pressurizer can be raised by operating electric heaters at the bottom of the pressurizer and can be lowered by introducing relatively cool water spray into the steam space at the top of the pressurizer.

The components in the lower end of the pressurizer (surge nozzle, lower head/heater well, and support skirt) are affected by pressure and surges through the spray nozzle. The components in the upper end of the pressurizer (spray nozzle, safety and relief nozzle, upper head/upper shell, manway and instrument nozzle) are affected by pressure, sprays through the spray nozzle, and steam temperature differences.

The limiting operating conditions of the pressurizer occur when the RCS pressure is high and the RCS hot leg (Thor) and cold leg (TCoId) temperatures are low. This maximizes the AT that is experienced by the pressurizer. Due to flow in and out of the pressurizer during various transients, the surge nozzle alternately sees water at the pressurizer temperature (Tsar) and water from the RCS hot leg at Thot. If the RCS pressure is high (which means, correspondingly, that Tsat is high) and Thot is low, then the surge nozzle will see maximum thermal gradients and thus experience the maximum thermal stress. Likewise, the spray nozzle and upper shell temperatures alternate between steam at Tsat and spray, which, for many transients, is at TCoId. Thus, if RCS pressure is high (Tsat is high) and Tcold is low, then the spray nozzle and upper shell will also experience the maximum thermal gradients and thermal stresses.

7.2 Key Input Assumptions The evaluation of the pressurizer used as input the NSSS design parameters and NSSS design transients applicable for the uprated conditions.

07/30/02 7-1

7.3 Acceptance Criteria The acceptance criteria for evaluating design inputs affecting the pressurizer stress report are as follows:

I. Hot- and cold-leg temperatures shall remain within the ranges of the operating temperatures that have previously been considered and justified in the pressurizer stress report.

2. NSSS design transients shall be less than or equal to the design transients previously considered in the pressurizer stress report with regard to both severity and numbers of occurrences.

3. Design loads shall be less than or equal in magnitude to the loads that were previously considered in the pressurizer stress report with no changes to the load application points and numbers of occurrences.

The acceptance criteria for the pressurizer structural analyses and evaluations are in accordance with the applicable requirements of the 1965 Edition of Section III of the ASME Boiler and Pressure Vessel Code with Addenda through the summer of 1966 (Reference 1) to which the PBNP pressurizer was originally designed.

7.4 Description of Analysis/Evaluation and Results The analysis performed for the PBNP pressurizer components to assess the impact of the thermal uprating conditions and 60-year life is based on the NSSS design transients. The transients were reviewed for conditions that differed from the conditions addressed in References 2, 3, and 4, to which the PBNP pressurizers were originally designed and analyzed. The thermal uprating include high and low primary temperature thermal-hydraulic conditions. To conservatively maximize thermal stresses (as described in Section 7.1), the lowest Thor and the lowest TCOId conditions were evaluated.

The analysis was performed by modifying the original PBNP stresses (References 2, 3, and 4) for the applicable pressure and thermal loads. Analytical models of the pressurizer components were subjected to pressure loads, external loads (such as piping loads), and thermal transients. The analysis includes calculations of the primary, secondary, and peak stresses for the various conditions.

The maximum pressure and maximum external loads on the pressurizer are not affected by thermal uprating conditions. Thus, the primary stresses calculated for the original analysis (References 2, 3, and 4) are still valid. The conditions that affect maximum primary-plus-secondary stresses do not change as a result of the thermal uprating, except for the surge nozzle. For the surge nozzle, the PL + Pb + Q stress intensity range is acceptable since it is less than the allowable stress limit of 3Sm = 57900 psi. For all the components, the fatigue analysis is not affected.

7-2 07/30/02 07/30/02 7-2

The original PBNP pressurizer surge nozzle analysis was previously updated for the thermal stratification pipe loads. The analysis update for thermal uprating considered all the previously reported changes to the original analysis.

The pressurizer component analysis also considered license extension to 60 years. The design transients indicate that the projected number of design transient occurrences for a 60-year design life are actually less than the original design value for a 40-year design life (based on plant operating data), except for the pressure test cycles. The pressurizer fatigue analysis assumed that the design transient occurrences for the 40-year design life are also applicable for the 60-year design. The revised pressure test cycles are incorporated in the analysis. Fatigue usage factors for the pressurizer components are provided in Table 7-1.

The results of the analysis indicate that the pressurizer components meet the ASME Code,Section III stress analysis and fatigue analysis requirements for the revised conditions and license extension transients, plus all other loadings given in the applicable design specifications. Analysis results also demonstrate that there is no impact on the primary stresses and the primary and secondary stresses (except for surge nozzle as mentioned above).

7.5 Conclusions The PBNP pressurizer components have been shown to meet the stress/fatigue analysis requirements of the ASME Code,Section III (Reference 1) using power uprate power conditions and the applicable design transients for 60 years of plant operation.

7.6 References

1. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda through Summer, 1966.

2. Wisconsin Electric Power Pressurizer Stress Report (for Point Beach Unit 1), Westinghouse Electric Corporation, Tampa Division, Tampa, Florida, August 1969.

3. Pressurizer Stress Report Sections 1 and 2, Wisconsin Electric Power Company, Point Beach Unit 2, Westinghouse Electric Corporation, Large Components Division, Tampa Division, Tampa, Florida, April 1975.

4. 51 Series Pressurizer Stress Report (Generic Stress Report), Sections 3.1 through 3.14, Westinghouse Electric Corporation, Tampa Division, Tampa, Florida, April 1975.

7-3 07/30/02 7-3

Table 7-1 PBNP Pressurizer Fatigue Usage Component Calculated Fatigue Usage Surge Nozzle [ ]b~c Spray Nozzle [ ]b,c Safety and Relief Nozzle [ ]bc Lower Head, Heater Well [ ]b~c Lower Head, Preformation [ ]b.C Upper Head and Shell [ ]bc Support Skirt/Flange [ ]b~c Manway Pad [ ]b,c Manway Cover [ ]b,c Manway Bolts [ ]b,c Support Lug ]bc Instrument Nozzle [ ]b,c Immersion Heater [ ]bc Valve Support Bracket [ ]bc Bracketed [ ] b,c information designates data that is Westinghouse Proprietary.

7-4 07/30/02 7-4

8.0 Auxiliary Components Evaluations were performed to assess the impact on the auxiliary components of a power uprate for PBNP Units I and 2 to a core power of 1650 MWt (NSSS power of 1656 MWt) for up to 60 years with renewed operating licenses.

8.1 Introduction The purpose of this evaluation is to determine the effect of the uprate conditions on the auxiliary equipment provided by Westinghouse, (i.e., auxiliary valves, pumps, tanks, and heat exchangers) for the following systems:

• Waste Disposal System

"* Sample System

"* Auxiliary Cooling System - including RHR, CCW and spent fuel clean-up

"* Chemical and Volume Control System - including the Boron Recycle System

"* Isolation Valve Seal Water

"* Feedwater System

"* Reactor Coolant System Specifically, the auxiliary valves, pumps and heat exchangers supplied by Westinghouse are listed in Tables 8-1, 8-2 and 8-3, respectively.

8.2 Key Input Assumptions The key input parameters for the evaluation were the following:

"* NSSS design parameters

"* Auxiliary design transients The original or equivalent technical and quality assurance requirements were used for any equipment that has been repaired or replaced.

8.2.1 Auxiliary Valves The original design parameters for the auxiliary valves were defined in the valve equipment specifications. The specifications for the individual valve types are listed in Table 8-1.

8-1 07130/02 8-1

The specifications contain the specific design parameters or reference the applicable valve specification sheet, which provides more detail as to the specific conditions for which the valve(s) were designed and manufactured.

8.2.2 Auxiliary Pumps, Heat Exchangers and Tanks The original design and manufacturing parameters for the equipment were defined in the design specifications and purchase orders. The purchase order and design specification for each individual pump type originally supplied for PBNP are listed in Table 8-2. The purchase order and design specification for each individual heat exchanger are listed in Table 8-3.

Various heat exchangers and tanks were supplied to PBNP; however, none of the tanks have significant transients specified as part of the design requirements.

The design specifications contain the general design, manufacturing, and qualification requirements. In addition to the design specifications, data sheets define specific requirements for each pump and heat exchanger.

8.3 Acceptance Criteria The primary acceptance criterion is that the auxiliary equipment transients, as applicable to the original design basis of the equipment remain unchanged or still bounding. If the transients are not affected, any original design analyses performed on the auxiliary equipment would remain applicable.

In addition, it has been assumed that any equipment maintenance or replacement was accomplished in accordance with the original equipment design requirements.

8.4 Description of Analysis/Evaluation and Results Auxiliary Valves The original design and qualification requirements for the auxiliary valves were reviewed. This was based on the requirements invoked by the equipment specifications, as identified in Table 8-1, applicable to PBNP. The specifications invoke transients and manufacturing requirements that influence the design analysis of the auxiliary valves. Normally, this review would involve a comparison of the specification requirements versus the revised transients and applicable manufacturing requirements to determine if any changes in the transients would have an effect on the existing auxiliary valve design analyses. Any changes would be reconciled by analyses. For this evaluation, however, the existing auxiliary system design transients bound the uprating conditions. Because the auxiliary equipment transients remained bounding, no auxiliary equipment valves are affected.

8-2 07/30/02 8-2

Auxiliary Pumps, Heat Exchangers and Tanks The design specifications define general equipment design, manufacturing, and qualification requirements, while the purchase orders and/or data sheets define specific design requirements. The original design, manufacturing, and qualification requirements for the auxiliary pumps listed in Table 8-2 and auxiliary heat exchangers listed in Table 8-3 were reviewed for potential impact by revised operating condition transients. Because the auxiliary equipment transients remain bounding, no auxiliary pumps or heat exchangers are affected.

As mentioned previously, none of the tanks have significant design transients specified. Therefore, the auxiliary tanks were not addressed further.

8.5 Conclusions Based on the review of the auxiliary equipment design conditions and design transients versus those for the uprating, the auxiliary pumps, valves, tanks, and heat exchangers supplied by Westinghouse are qualified for the Point Beach Units 1 and 2 at the uprated core power of 1650 MWt. Therefore, the original design analyses remain applicable for the qualification of the auxiliary mechanical equipment.

Further, since the auxiliary system design transients took into account 60 years of operation, this qualification of the equipment applies for that same length of time.

8-3 07/30/02 07/30/02 8-3

Table 8-1 PBNP Auxiliary Valves Number/Revision Specification Title G-67624 1/1 Manual T & Y Globe, Manual Gate, and Self-Actuated Check Valves 676270/1 Control Valves 676281/2 Diaphragm-Type Valves 676258/2 Motor-Operated Valves 676368/1 Butterfly Valves G-676257/2 Auxiliary Relief Valves 676279/2 Pressurizer Safety Valves 677264/0 2 in. and below Manual T & Y Globe, and Self-Actuated Check Valves 955712/0 1/8 thru 3/8 Manually-Operated Globe, Throttle, and Check Valves.

ASME Boiler and Pressure Vessel Code Section III, Class 2 And 3.

G-952955/0 1/8 thru 3/8 Manually-Operated Globe, with Interim Change 1 Throttle and Check Valves. ASME Boiler and Pressure Vessel Code Section III, Class 2 and 3.

8-4 07/30/02 8-4

Table 8-2 PBNP Auxiliary Pumps W Purchase Specification Pump Description Order No./Rev. Vendor Spent Fuel Pool Pump 66168 676428, Rev. 0 Ingersoll-Rand Boric Acid Recirc Pump 134809 677123, Rev. 0 Ingersoll-Rand Containment Spray Pump 66168 676428, Rev. 0 Ingersoll-Rand Component Cooling Water 70491 676428, Rev. 0 Ingersoll-Rand Sump Tank Pump 67798 676428, Rev. 0 Goulds Monitor Tank Pump 67798 676428, Rev. 0 Goulds Waste Condensate Pump 67798 676428, Rev. 0 Goulds Laundry & Hot Shower 67798 676428, Rev. 0 Goulds Waste Evaporator Feed Pump 67798 676428, Rev. 0 Goulds Refueling Water Purification 67798 676428, Rev. 0 Goulds Chem Drain Tank Pump 67798 676428, Rev. 0 Goulds Residual Heat Removal 70504 676428, Rev. 0 Pacific Pumps WGS Compressor Package 70518 676451, Rev. 1 Nash Gas Stripper Feed Pump 72402 676428, Rev. 0 Chempump RCS Drain Tank Pump 72402 676428, Rev. 0 Chempump Conc Holding Tank Transfer 72402 676428, Rev. 0 Chempump Boric Acid Transfer Pump 72402 676428, Rev. 0 Chempump CVCS Charging Pump 65265 676326, Rev. 0 Ajax Iron Works Safety Injection Pump 77771 676428, Rev. 0 Byron Jackson Spent Fuel Pool Skimmer 86068 676428, Rev. 0 Duriron 8-5 07/30/02 07/30/02 8-5

Table 8-3 PBNP Auxiliary Heat Exchangers Spin Number Data Sheet Equipment Component/Manufacturer Drawing No./Rev. Number Specification P.O. Number Regenerative Heat Exchanger CSAHRG AH-RG503 G-676454 54-Z-66179-B Sentry Equipment Corp. A04195-AO1-2, Rev. 8 Rev. 0 Residual Heat Exchanger ACAHRS Sheet B G-676228 59-Z-70134-B Joseph Oat & Sons, Inc. 4836, Rev. 2 Rev. 0 Component Cooling Heat Exchgr ACAHCC Sheet D G-676228 54-Z-70135-B Atlas Industrial Mfg. Co. D-1260, Rev. 4 dated 1/5/67 Rev. 0 Seal Water Heat Exchanger CSAHSW Sheet C G-676228 54-Z-70135-B Atlas Industrial Mfg. Co. D-1259, Rev. 3 dated 1/5/67 Rev. 0 Excess Letdown Heat Exchanger CSAHEL Sheet E G-676228 54-Z-66179-B Sentry Equipment Corp. P-6703-4-1, Rev. 9 dated 1/5/67 Rev. 0 Sample Heat Exchanger SSAHSS Sheet G G-676228 54-2-66179-B Sentry Equipment Corp. P-6780-2-1, Rev. 6 dated 1/5/67 Rev. 0 Spent Fuel Pool Heat Exchanger ACAHCC Sheet F G-676228 54-Z-70135-B Atlas Industrial Mfg. Co. D-1253, Rev. 3A or 4(1) Rev. 0 Non-Regenerative Heat Exchgr SSAHSS Sheet J-2X G-676228 54-2-66179-B Sentry Equipment Corp. A03996-54-1, Rev. 2 Rev. 0 Note:

1. The Rev, 4 and Rev. 3A drawings are the same drawing, but it is called by both names in various places.

07/30/02 8-6

9.0 Reactor Coolant System Supports Evaluations were performed to assess the impact on the reactor coolant system (RCS) supports of a power uprate for PBNP Units I and 2 to a core power of 1650 MWt fori up to 60 years with renewed operating licenses.

9.1 Introduction The uprate RCS parameters associated with the power uprate were reviewed for potential impact on the primary equipment supports. The revised temperatures may cause potential load changes in the components to be reconciled. The thermal design transients applicable to the uprate were also factored into the evaluation. The RCS equipment evaluated includes:

"* Steam Generator (SG) Lower Lateral Frame

"* SG Columns

"* SG Upper Supports

"* RCP Tie Rods

"* RCP Columns

"* Reactor Vessel (RV) Support 9.2 Key Input Parameters The evaluation assumes that all criteria and methods used in the existing design basis for PBNP will continue to be used.

Four basic sets of input parameters are used in the evaluation of the RCS supports listed above:

• NSSS Performance Parameters

• Thermal Design Transients

• Piping Loads

• Plant Life Extension to 60 years 9.3 Acceptance Criteria The acceptance criteria for structural steel associated with primary equipment supports follows PBNP FSAR Appendix A, supplemented by the American Institute of Steel Construction (AISC) "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" (Reference 1) and the ASME subsection NE (Reference 2). The acceptance criteria for component supports as stated in Table A.5-3 of 07/30/02 9-1

the PBNP FSAR, provides stress limits and specific values for defined loading conditions. Table 9-1 states the acceptance criteria as defined in the PBNP FSAR.

9.4 Description of Analysis/Evaluation and Results An evaluation was performed of the current piping supports analysis. The STRUDL and WESAN computer codes (Reference 3 and 4 respectively) have been used for the structural support analysis since the original plant design.

The primary equipment supports were not a Westinghouse design and many of the design basis calculations were not available. The primary equipment support assessment consisted of reviewing the support loadings and comparing the calculated combined loads with the allowable loads for "upset" and "faulted" conditions. Seismic loads are not combined with pipe rupture loads. Large loop break (LOCA) loads have been eliminated by the application of leak-before-break (LBB) methodology as discussed in Section 10. The allowable stress criteria used in the assessment followed the PBNP FSAR Appendix A to the extent possible and was supplemented by other AISC or ASME NF criteria as required.

The evaluation to 1650 MWt showed that all the supports are within allowable limits and thus structurally adequate. See Tables 9-1 and 9-2 for the results of the analyses and evaluations.

9.5 Conclusions The design parameters for the core power uprate to 1650 MWt (1656 MWt NSSS power) have been evaluated for impact on the primary equipment supports. The evaluation indicates that all components meet appropriate allowables. The evaluation for the stated components concludes that there is no adverse effect on the ability of these components to operate until the scheduled end of plant operation (60 years).

9.6 References I. AISC "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,"

Eighth Edition.

2. ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, entitled Component Supports, 1974 Edition.

3. STRUDL- Structural Analysis Computer Code, Version Date 1978.

4. WESAN - Westinghouse Electric Support Analysis, Version Date 1979.

9-2 07/30/02 9-2

Table 9-1 Support Structure Loading Combinations and Stress Criteria For Normal, Upset, and Faulted Conditions Stress Limit Operating Plate and Shell Condition Loading Combination Linear-Type Supports Supports Normal Thermal expansion, Weight, Within working limits Pm < Sm Operating pressure Pm + Pb < 1.5 Sm Upset Thermal expansion, Weight, Within working limits N/A Operating pressure OBE Faulted Thermal expansion, Weight, Yield Stress Limits N/A Operating pressure, Weight, ASME NF criteria LOCA/Pipe Rupture, DBE 9-3 07/30/02 07/30/02 9-3

Table 9-2 Primary Equipment Support Member Stresses Member Stresses (Percent of Allowable)

Member Upset Faulted SG Lower Lateral Frame [ b,c [ b,c SG Columns [ ]b,c b,c SG Upper Supports _ ]bc RCP Tie Rods (2) [ ]bc RCP Columns [ ]bC [ ]b,c RV Support [ ]bC [ ]b,c Notes:

I. The SG upper support loads are 1/2 the DBE loads.

2. There are no loads on the RCP tie rods due to a 3/8" hot gap.

Bracketed [ ] b,c information designates data that is Westinghouse proprietary.

07130/02 9-4 07/30/02 9-4

10.0 Application of Leak-Before-Break Methodology 10.1 Introduction The original structural design basis of the reactor coolant system (RCS) for PBNP Units 1 and 2 required that dynamic effects resulting from pipe breaks be considered and that protective measures for such breaks be incorporated into the design. Subsequent to the original PBNP design, the additional concern of asymmetric blowdown loads was raised as described in Unresolved Safety Issue A-2 (Asymmetric Blowdown Loads on the Reactor Coolant System) and Generic Letter 84-04 (Reference 1). However, research by the Nuclear Regulatory Commission (NRC) and industry, coupled with operating experience, determined that safety could be negatively impacted by placement of pipe whip restraints on certain systems. As a result, NRC and industry initiatives demonstrated that leak-before-break (LBB) criteria can be applied to RCS piping based on fracture mechanics technology and material toughness.

Generic analyses for the application of LBB for specific plants was documented in response to Unresolved Safety Issue A-2 and approved for PBNP in an NRC letter dated May 6, 1986 (Reference 2).

In that letter, the NRC stated that PBNP's request for an exemption from the requirements of 10 CFR 50 Appendix A, General Design Criterion (GDC) 4, to eliminate the consideration of large RCS primary loop pipe breaks in the PBNP structural design basis was not necessary since the Commission had published its final rule in the Federal Register on April 11, 1986 (Reference 3). This change to the rule allowed the use of LBB technology for excluding from the design basis the dynamic effects of postulated ruptures in primary coolant loop piping in pressurized water reactors.

This section describes the LBB analysis of the PBNP primary loop piping, which was evaluated for a core power level of 1650 MWt.

10.2 Key Input Parameters The parameters that are important in the analysis are the piping forces and moments and normal operating temperature and pressure. These parameters are required in the determination of leakage flaw size and critical flaw size. They are also required in the crack stability analysis.

The LBB evaluation for the primary loop piping was performed based on conditions associated with NSSS design parameters. The loadings for the primary loop piping were used in the evaluation.

10.3 Acceptance Criteria In order to demonstrate the elimination of breaks in the RCS primary loop pipe for the PBNP plants, the following objectives must be achieved:

Demonstrate that margin exists between the critical crack size and a postulated crack which yields a detectable leak rate.

07/30/02 10-1

"* Demonstrate that there is sufficient margin between the leakage through a postulated crack and the leak detection capability of the PBNP plants.

"* Demonstrate margin on applied load.

"* Demonstrate that fatigue crack growth is negligible.

The LBB acceptance is based on the Standard Review Plan (SRP) 3.6.3 (Reference 4). The recommended margins are as follows:

"* Margin of 2 on Flaw Size

"* Margin of 10 on the Leak Rate Margin of >1 on Loads (using faulted load combination by absolute summation method).

10.4 Description of Analysis/Evaluation and Results Primary Loop Piping The current LBB analysis for the PBNP Units 1 and 2 primary loop piping incorporated a core power uprate to 1650 MWt. The results of the analysis were documented in WCAP-14439 (Reference 5). The report demonstrated compliance with LBB technology for the PBNP primary loop piping based on plant specific analysis. The margins determined in Reference 5 are described below.

Flaw Size - Using faulted loads obtained by the absolute sum method, a margin of 2 exists between the critical flaw and the flaw having a leak rate of 10 gpm (the leakage flaw).

Leak Rate - A margin of 10 exists between the calculated leak rate from the leakage flaw and the leak detection capability of 1 gpm.

Loads - A margin of > 1 on loads exists by using the absolute sum method of faulted loads combination.

The evaluation showed that all the LBB recommended margins documented in Reference 5 remain valid for the uprate condition of 1650 MWt core power (Reference 5).

10.5 Conclusions The LBB criteria are satisfied for the PBNP Units 1 and 2 primary loop for a core power uprate. All the recommended margins are satisfied. It is therefore concluded that the dynamic effects of the RCS pipe breaks analyzed need not be considered in the structural design basis of the PBNP Units 1 and 2 for a core power uprate to 1650 MWt.

10-2 07/30/02 07/30/02 10-2

10.6 References

1. USNRC Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1, 1984.

2. NRC Docket Nos. 50-266 and 50-301 Letter from G. E. Lear, Director PWR Project Directorate No. 1 Division of PWR Licensing-A, NRC, to C. W. Fay, Vice President Nuclear Power Department Wisconsin Electric Power Company.

3. NRC 10 CFR 50, "Modification of General Design Criteria 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures," Final Rule, Federal Register/Vol. 52, No. 207/Tuesday, October 27, 1987/Rules and Regulations, pp. 41288-41295.

4. Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday August 28, 1987/Notices, pp. 32626-32633.

5. WCAP-14439 Revision 0, "Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2 For The Uprate Program," February 1996, Westinghouse Proprietary Class 2.

10-3 07/30/02 07/30/02 10-3

RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 226 MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Attachment 8 FSAR Markups for 14.1.9, "Loss of External Electrical Load," 14.1.10, "Loss of Normal Feedwater," and 14.1.11, "Loss of All AC Power to the Station Auxiliaries" (for information only)

PBl'P FSAR (06/01) Loss of External Electrical Load Page 14.1.9-1 of I I 14.1.9 Loss of External Electrical Load The plant is designed to accept a 50% loss of electrical load while operating at full power or a complete loss of load while operating below 50% power without actuating a reactor trip. The automatic steam bypass system with 40% steam dump capacity to the condenser is able to accommodate this load rejection by reducing the transient imposed upon the reactor coolant system. The reactor power is reduced to the new equilibrium power level at a rate consistent with the capability of the rod control system. Should the reactor suffer a complete loss of load from full power, the reactor protection system would automatically actuate a reactor trip.

The most likely source of a complete loss of load'on the nuclear steam supply system is a trip of the turbine-generator. In this case, there is a direct reactor trip signal derived from either the turbine autostop oil pressure or a closure of the turbine stop valves, provided the reactor is operating above 50% power. Reactor temperature and pressure do not increase significantly if the steam bypass system and pressurizer pressure control system are functioning properly.

However, the plant behavior is evaluated for a complete loss of load from full power without a direct reactor trip, primarily to show the adequacy of the pressure relieving devices and also to show that no core damage occurs. The reactor coolant system and steam system pressure relieving capacities are designed to ensure the safety of the plant without requiring the automatic rod control, pressurizer pressure control, and/or steam bypass control systems.

Method of Analysis The total loss of load transients are analyzed by employing the detailed digital computer program LOFTRAN (Ref. 2). The program simulates the neutron kinetics, reactor coolant system,. pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves.

The program computes pertinent plant variables, including temperatures, pressures, and power level.

In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from full power without direct reactor trip, primarily to show the adequacy of the pressure-relieving devices and also to demonstrate core protection margins.

• C*(Re--t),Plant characteristics and initial conditions are discsussed in Section 14.

SInitial- Oertig onitons h ntial core powe-r,reactor coolant tenmp ature, a~nd reactor coolant pesr r sue ttems iiignmný aus h Rcluain r perform gn Proc in whc the uncertainisithintacodtosaeicueinteDBliivau.FrtepkRC pressure calculations, uncertainties of 2%, 50 psi, and 6'F are applied in the most limiting direction to the initial core power, reactor coolant pressure, and reactor coolant temperature.

t,.-,* ;,1 t ,, *,

• 4 -  :- . . . ....- ,

PBNP FSAR (06/0 1) Loss of Extmnal Electrical Load Page 14.1.9-3 of I I PBNPFSAR (06/01) Loss of External Electrical Load Page 14.1.93 of II Figure 14.1.9-2 shows the total loss of load accident, assuming the plant to be initially operating at full power, with no credit taken for the pressurizer spray, pressurizer power-operated relief valves, or steam dump. The reactor is tripped on the high pressurizer pressure signal. In this case, the pressurizer safety valves are actuated.

The calculated sequence of events for these two cases is shown in Table 14.1.9-1.

Conclusions Results of the analyses show that the plant design is such that a total loss of external electrical load without a direct or immediate reactor trip presents no hazard to the integrity of the reactor coolant system or the main steam system. Pressure-relieving devices incorporated in the two systems are adequate to limit the maximum pressures within the design limits.

The integrity of the core is maintained by operation of the reactor protection system; i.e., the departure from nucleate boiling ratio is maintained above the limit value.

References 6j_. ,c-,* .

1. Friedland, A. J., Ray S., "Revised Thermal Design Procedure," WCAP- 1139

2. Burnett, W. E., et. al., "LOFTRAN Code Description," WCAP-7907,4O 97

//

PBNýP FS AR (06/01 ) Loss of External Electrical Load Page 14.1.9-4 of I I PBNP FSAR (06/01) Loss of External Electrical Load Pace 14 1 9-4 of II TABLE 14.1.9-1 TIME SEQUENCE OF EVENTS FOR LOSS OF EXTERNAL ELECTRICAL LOAD Case Event Time of Each Event (Seconds)

a. With pressurizer control Loss of electrical load 0 (minimum eedbck)Overtemperature AT reactor feedback)

(minmum

/ "-Initiation of release

' / from SG safety valves Rod begins to drop k33 Peak RCS pressure occurs Minimum departure from nucleate boiling ratio occurs

b. Without pressurizerc Loss of electrical lod (minimum feedback)

High pressurizer pressure reactor trip point reached 6.4 Rods begin to drop 8.4 Peak RCS pressure occurs 9.4 Initiation of release from SG safety valves 9.7

PBNP FSAR (06/01) Loss of External Electrical Load Page 14.1.9-5of 11 TABLE 14.1.9-2 MSSV CHAR.ACTERISTICS Paramet Bank 1 Bank 2 Bank 3 ank4 Nominal set p ssure (psia) 1100 1115 1140 1140 Lift pressure (psia) 1153 1169 1 5 1195 Relief rate at life pressur 0 0 0 0 (lbm/sec per valve)

Full-open pressure (psia) 1165.1 11 .7 1206.7 1206.7 Relief rate at full-open 230 230 230 230 pressure (lbm/sec per valve)

Table 14.1.9-2 Notes

1. The lift pressure is the noy~nal set pressure, plus %7allowance for setpoint tolerance, plus 20 psi allowance f frictional pressure drop be een the steam -enerator shell and the valve.

2. The full-open pr sure is the lift pressure, plus 1.1% allow ce for valve accumulation (1.1% of the minal set pressure).

3. The MSS relief rate is assumed to be a linear function of the pre ure between the lift pre ure and the full-open pressure.

4 V es for the lift pressure and full-open pressure listed above reflect the G shell ressure, which will be greater than the steam pressure at the safety valve e to the frictional pressure drop from the SG shell to the safety valves (assumed to be 0 psi).

These values should be reduced by the frictional pressure drop to obtain the act I setpoints supported by the analysis.

TABLE 14.1.9-2 MSSV CHARACTERISTICS Parameter Bank I Bank 2 Bank 3 Bank 4 Nominal set pressure (psig) 1085 1100 1125 1125 Lift pressure (psia) 1132.3 1148.8 1177.7 1177.7 Relief rate at lift pressure 0 0 0 0 (Ibm/sec per valve)

Full-open pressure (psia) 1178.6 1197.4 1236.2 1236.2 Relief rate at full-open 230 230 230 230 pressure (Ibm/sec per valve)

Table 14.1.9-2 Notes

1. The lift pressure is the nominal set pressure, plus 3% allowance for setpoint tolerance (3% of the nominal set pressure), plus the appropriate allowance for the frictional pressure drop between the steam generator shell and the valve at the lift pressure, plus atmospheric pressure (14.7 psi).

2. The full-open pressure is the lift pressure, plus 1.1% allowance for valve accumulation (1.1% of the sum of the nominal set pressure and tolerance), plus the appropriate allowance for the frictional pressure drop between the steam generator shell and the valve at the full-open pressure, plus atmospheric pressure.

3. The MSSV relief rate is assumed to be a linear function of the pressure between the lift pressure and the full-open pressure.

4. The values listed above for the lift pressure and full-open pressure reflect the SG shell pressure. However, since the safety valves are actually located downstream of the SG, the pressure at the valve is only the same as that in the SG when the first safety valve opens. Once relief flow is established, a frictional pressure drop will exist between the SG shell and the valves (assumed to be 50 psi at full relief flow) and the steam pressure at the safety valve will actually be less than the values listed above. Thus, the appropriate allowance for the frictional pressure drop has been conservatively included in the values listed above for the lift pressure and full-open pressure.

PBNP FSAR (06/01) Loss of External Electrical Load Page 14.1.9-6 of 11 FIGURE 14.1.9-1 Sheet I of 3 LOSS OF ELECTRICAL LOAD WITH PRESSURE CONTROL MINI-MUM REACTIVITY FEEDBACK

1.,4 E

0 f- 1 0

0 .4

.2

3 0 o .4 0 20 40 so 80 100 120 Time (seconds) 660 "Co 0 S6Z Eooo 0 06 540 Io 108 1 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Loss of Electrical Load With Pressure Control Minimum Reactivity Feedback Figure 14.1.9-1 (sheet 1 of 3)

PBNP FSAR (06/01) Loss of External Electrical Load Page 14.I.9-7 of I I PBNP ESAR (06/01) Loss of External Electrical Load Page 14.19-7 of II FIGURE 14.1.9-1 Sheet 2 of 3 LOSS OF ELECTRICAL LOAD WITH PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK LAJCx t NF

---

I

• "22o00 a

S2400 CL 0 240 60 30 1O0 IN Time (seconds) 800 S700 E

D 6000 0

- 5.w 0400 D

S 40 80 I30 120 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Loss of Electrical Load With Pressure Control Minimum Reactivlty Feedback Figure 14.1.9-1 (sheet 2 of 3)

PBrNT FSAR (06/01) Loss of External Electrical Load Page 14.1.9-8 of I11 PBNT FSAR (06/01) Loss of External Electrical Load Page 14.1.9-SoflI FIGURE 14.1.9-1 Sheet 3 of 3 LOSS OF ELECTRICAL LOAD WITH PRESSURE CONTROL MINIMUM REACTIVITY FEEDBACK R ct IWO I F 13 <

1300 1200D S1100

.- 1000 7D0O

• . 700 V) 0 20 40 60 I0 100 120 Time (seconds)

.5 2.5.

Point Beach Nuclear Plant Units 1 and 2 Loss of Electrical Load With Pressure Control Minimum Reoctivity Feedback Figure 14.1.9-1 (sheet 3 of 3)

PBNP FSAR (106/01 ) Loss of Normal Feedwater Page 14.1. 10- 2 of 16 Method of Analysis A detailed analysis using the LOFTRAN code is performed in order to obtain the plant transient following a loss of normal feedwater. The simulation describes the plant thermal kinetics, RCS including the natural circulation, pressurizer, steam generators. and feedwater system. The digital program computes pertinent variables, including the steam generator level, pressurizer water level, and reactor coolant average temperature.

The following assumptions were made:

1. The plant is initially operating at 102% of 1524.5 MWt.

2. Core residual heat generation is based on the 1979 version of ANS-5.1 (Reference 1) plus two standard deviations for uncertainty. ANSI/ ANS-5.1-1979 is a conservative representation of the decay heat release rates.

3. One motor driven auxiliary feedwater pump provides 200 gpm of flow split to two steam generators, 5 minutes following receipt of a low-low steam generator water level setpoint signal.

4. The assumed steam generator models are 44F (Unit 1) and Delta-47 (Unit 2).

5. The pressurizer sprays, heaters, and PORV's function to produce the maximum peak pressurizer water volume.

Results The calculated sequence of events for this event is listed in Table 14.1.10-1. Figures 14.1.10-1 and 14.1.10-2 show the plant parameters following a loss of normal feedwater accident with the assumptions listed above for Units 1 and 2. Low-low level signal in either steam generator initiates the reactor trip. The reactor trip then initiates the turbine trip. Following the reactor and turbine trip from full load, the water level in the steam generators falls due to the reduction of steam generator void fraction and because steam flow through the safety valves continues to dissipate the stored and generated heat.

Upon the initiation of the low-low level signal, the auxiliary feedwater pumps are automatically started. The pumps will supply auxiliary feedwater to one steam generator within five minutes, reducing the rate of water level decrease.

The capacity of the auxiliary feedwater system is such that the water level in the steam generators does not recede below the-lowest level at which sufficient heat transfer area is available to dissipate core residual heat without water relief from the RCS relief or safety valves. From Figures 14.1.10-1 and 14.1.10-2 it can be seen that at no time is there water relief from the pressurizer.

G.Tke MSsV 0c,I cskec t~pc recri '. W

PBNP FSAR (06/01) Loss of Normal Fcedwater Paoe 14.1.10-4 of 36 PBN? FSAR (06/01) Loss of Normal Feedwater Pate 14.1.10-4 of 16 TABLE 14.1.10-1 TLME SEQUENCE OF EVENTS FOR LOSS OF rNORMAL FEEDWATER FLOW INCIDENTS Time of Each Event (Seconds)

Event UnitI Main feedwater flow stops 10 1C C Low-Lfow steaam goernerrator water level trip 46 9~ 7.5 Rods begin to drop 4.

Two steam generators begin to receive auxiliary 334 . 34 feedw ater Cold auxiliary feedwater is delivered to the steam 435.0 437.

generators Peak water level in pressurizer occurs 3 0 3>

Core decay heat decreases to auxiliary feedwater heat < 6000 < 60(

removal capacity

PBNP FSAR (06/01) Loss of Normnal Feedwater Page 14.1.10-5 of 16 FIGURE 14.1.10-1 Sheet 1 of 6 Rcrio~cc AJ

E 0

C:

0 0 .6 0

Time (seconds) o .2 0

"C 0 *213 4 010 . 10.. 0 ID 10 "Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 4-4F SG)

Loss of Normal Feedwoter Figure 14.1.10-1 (sheet I of 6)

PBNP FSAR (06/01) Loss of Normal Feedwater Page 14.1.10-6 of 16 FIGURE 14.1.10-1 Sheet 2 of 6 Time oln 2\

Unit 1 (Model 4-4F Sý Lozz of Normal Feedw, Fgure 14.110-1 (1heet 2 of 6) p,c +/-ic, F v;--

Reactor Vessel Flow (fraction of Initial) Core Reactivity (pcm) 61 6,i I

• I ' : T I , , I I . q Ir * . I I I a

,4 0

0 0 o4 a.

(L 8-- (A ;S_-

0° r

0 ZJ (4

.5 °-,

a. a 0.

o3 0)0 oCD 0

PBNP FSAR (06/01) Loss of Normal Feedwater Page 14.1.10-7 of 16 FIGURE 14.1.10-1 Sheet 3 of 6 iD Time Nexk

700 660 0

620 0

I 500 o

o 0

540

-o 0

S* : * ' . .: I , I C . .I = ' ' . .

06 10 10, 10 10 10 Time (seconds)

I 700 I

cx 660 620 I

o_

o 5400

-o 0J 16 ID 10 3

10 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Mlodel 44F SG)

Loss of Normal Feedwoter Figure 14.1.10-1 (sheet 3 of 6)

PBNP FSAR (06101) Loss of Normal Feedwater Page 14.1-10-8 of 16 FIGURE 14.1.10-1 Sheet 4 of 6 1000 E

IO

'C-.*._.

C) 10 Tm Time N, t Fi 3 r

2300 0

2200 2100 2000

¢D 1900 18DO I

I-1700 l~n . ' : * . . .. T . . . I . . . . . . . .

I 10 10 l0 10 Time (seconds)

E 0

i>.0 Co No a-Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 4-4F SG)

Loss of Normal Feedwater Figure 14.1.10-1 (sheet 4 of 6)

lid 0

(31 0

0

~T 0~

I CL Q1 C.)

0 1200 -I Ic DO4 E

o 0

0

-3

+/-0 S10, 10, 10 10 10, Time (seconds) 1200

3_ +/-i CL E

C'4 a* 600 0

0

-1 400 10 ID 10 10 10 10 lime (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 44F SG)

Loss of Normal Feedwater FIgure 14.1.10-1 (sheet 5 of 6)

PBNP PSAR (06/01) Loss of Normal Feedwater Pa~ge 14. 1.10-10of 16 FIGURE 14.1.10-1 Sheet 6 of 6 100000 E

00000 0

0 500 0 0 0

400 00 Z000 0 0

0 0

-J 10 T-ime X .

in 2ý Unff 1 (Model 44FWS Lo-= of lNorrnat Feedwo r F-Igur-e 14-1.10-1 (a1heet 6 of 6)

Nc)(* F~i 3 (,cce

-o C

CD 0

2EOOOO 0 o 0

0 1010, 10 10 10 "T"me (seconds) 10M00 i 000000

-4w*

0 C

0 0

0 . . . . . . i1 . . . . . . .13 . . .

10, ID 10, 10 10 10 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unif 1 (Model 44F SG)

Loss of Normal Feedwafer Figure 14.1.10-1 (sheet 6 of 6)

PBNP FSAR (06/01) Loss of Normal Feedwater Page 14.1.10-11 of 16 PBNP ESAR (06/01) Loss of Normal Feed water Page 14.1.10-11 of 16 FIGURE 14.1.10-2 Sheet I of 6 Time Point Beach Nu>

Unifs 1 01 Unit 2 (Delt--47 Sd Lo= of Normal Feedw, Figure 14.1.10--2

,3heet 1 of 6)

Rz 1o'ce" wkJh Y

N F ,19 L,-r?

Core Heat Flux (fraction of nominal) Nuclear Power (fraction of nominal) 3_

CD to 0 0 0 0 a

to, 0

0 toC o Cto:

(D 2b. .0 2::

0.

"70

-D

-u

PBNP FSAR (06/01) Loss of Normal Feedwater Page 14. 1. 1G- 12 of 16 FIGURE 14.1.10-2 Sheet 2 of 6 Rcpla~ct. vJ+k

1000 0

E 4

0 -000

-5000 1010 10 10 10 10 Time (seconds)

.2

,Z3 0

0=

.9

.2 01: 10 10 to to Time (seconds)

Point Beach Nuclear Plant Units I and 2 Unit 2 (Della-47 SG)

Loss of Normal Feedwofer Figure 14.1.10-2 ksheer 2 of 6)

PBNP FSAR (06/01) Loss of Normal Feedwater Page 14.1.10-13 of 16 FIGURE 14.1.10-2 Sheet 3 of 6 10 lime Rcp['c' Iijkk*

D~

I 700 I CD I I ----------------------------------I--------------------------

" 62O+

0 I 080

-1, I-0 0I.

61 10I 10 10 10 Time (seconds)

I

~0

"-6_-o 0

0 0

-0 0

32 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 2 (Delta-47 SG)

Loss of Normal Feedwoter Figure 14.1.10-2 (sheet 3 of 6)

.1 ______________________________________________________________________________

PBN'P FSAR (06/01) Loss of Normal Feedwater Pagec 14.1.10-14 of 16 PBNP FSAR (06/0]) Loss of Normal Feedwater Page 141.10-14 of 16 FIGURE 14.1.10-2 Sheet 4 of 6 N0 0**

N,-

1T0 Time

2400 7300 2100 EL PD 4-1700-10 10 10 10 10 1 Time (seconds)

(

N0 0.

Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 2 (Delta-47 SG)

Loss of Normal Feedwater Figure 14.1.10-2 (sheet 4 of 6)

I _______________________________________

PBN'P FSAR (06/01) Loss of Normal Feedwater Page 14. 1. 15 of 16 PBNP ESAR (06/01) Loss of Normal Feedwater Page 14110-15 of 16 FIGURE 14-1.10-2 Sheet 5 of 6 10 10 Time Point Beach Nulear Plant Units I anf 2 Unit 2 (Defto-47 St Loe of Normal Fesdwi Flgure 14.1.10-2

(:heet 5 of 6)

Rceltacc UJH-k Nce4,tF .f

14WA ,

0 I

=004 E

0~

in Bw 4 0 600w 0

-J I 3 Il 10 10 s 0 10 Time (seconds) 0.

E 0

on Co4 0L Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 2 (Delta-47 SG)

Loss of Normal Feedwater Figure 14.1.10-2 (sheet 5 of 6)

Loop 2 SG Mass Inventory (Ibm) 0 o C ,m 0 o o 0 ,o o m

0 0.

A 0N-

0 CD 0.

2oc

0. 200 10 10

,0 0 10 10 10 Time (seconds) 1o0 o E

~ov Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 2 (Delta-47 SG)

Loss of Normal Feedweier Figure 14.1.10-2 (sheet 6 of 6)

PBNTP FSAR (06/01) Loss of All AC Power to the Station Auxiliaries Page 14.'. 11 -2 of15 PBNP PSAR (06/01) Loss of All AC Power to the Station Auxiliaries Pace 14.1.11-2 of 15 The following assumptions are made:

1. The plant is initially operating at 102% of 1656 MWt.

2. Core residual heat generation is based on the 1979 version of A.NS-5.1 (Reference 1) plus two standard deviations for uncertainty. ANSI/ANS-5.1 - 1979 is a conservative representation of the decay heat release rates.

3. The AFW system provides 200 gpm of flow to one steam generator, 5 minutes following receipt of a low-low steam generator water level setpoint signal.

4. The assumed steam aenerator models are 44F (Unit 1) and Delta-47 (Unit 2).

5. Secondary system steam relief through the self-actuated safety valves.

6. After normal steam generator level is established, auxiliary feedwater flow is controlled to maintain the water level.

7. The pressurizer sprays, heaters, and PORV's are assumed to function as designed which maximizes the peak pressurizer water volume.

The remaining assumptions used in the analysis are similar to the loss of normal feedwater (14.1.10) except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip plus an appropriate delay time (2 sec. for reactor trip and 2 sec. for loss of power for a total of 4 sec.).

Results The calculated sequence of events for this accident is listed in Table 14.1.11-1. The transient response of the RCS following a loss of AC power is shown in Figures 14.1.11-1 and 14.1.11-2.

The first few seconds after the loss of power to the reactor coolant pumps will closely resemble the simulation of the loss of reactor coolant flow event (14.1.8), where core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor.

After the reactor trip, stored and residual decay heat must be removed to prevent damage to either the RCS or the core.

The results of the analysis show that the natural circulation flow available is sufficient to provide adequate core decay heat removal following reactor trip and RCP coastdown.

Conclusion The loss of AC power to the station auxiliaries does not cause any adverse condition in the core, since it does not result in water relief from the pressurizer relief or safety valves.

References

1. "American National Standard for Decay Heat Power in Light Water Reactors,"

ANSIIANS-5.1 - 1979, August 1979.

2. Burnett, T. W. T., et.al., "LOFTRAN Code Description." WCAP-7907-P-A (Promrietar- ', CAP-7907-A (non-Propri April 1984 (gree i1. >,i.SV -&r G czicc-ckcc4vccJ

PB NP FS AR (0610 1) Loss of All AC Power to the Station Auxiliaries Pag2e 14. 1.11 -3 o f 15 PBNP PSAR (06/Oh Loss of All AC Power to the Staiion Auxiliaries Pa2e 4 I 11 of I TABLE 14.1.11-1 TIME SEQUENCE OF EVENTS FOR LOSS OF OFFSITE POWER ENCDDENTS*

Time of Each Event (Seconds)

Event Unit 1 Unit 2 Main feedwater flow stops 10 10 Low-Low steam generator water level trip 159 Rods begin to drop 4

) Reactor coolant pumps begin to coastdown - L One steam generator begins to receive auxiliary feedwater Cold auxiliary feedwater is -, -7 0 delivered to the steam generator

/

/

Peak water level in pressurizer occurs K ,

z Core decay heat decreases to auxiliary feedwater heat removal capacity < 2500 :0-'

T

• N onem ergency AC power to station auxiliaries is lost at o ...- .. .....

-4 se.v.uia

• 2 /n

PBNP FSAR (06/01) Loss of All AC Power to thec Station Auxiliaries Page 14. 1.11--'of 15 FIGURE 14.1.11-1 Sheet I of 6 "UNIT I (MODEL 44F SG) LOSS OF AC POWER C

C

-7 0

C C

U C

,-"\

C C

I

• 3 ,i to Time (sconds)

N C

0 L

. 1 0 L 1

C C

L C

L 0

C.)

IQI TTme (seceonds)

Point Beach Nutisar PIant Units I on8 2 UnvIt I (hiod.4 44F SGý.

Loss of AC Power tet1 of 6)

I

\

52 C

C E

0 0

1-2 1.2 0

E 50 1.0 1. 10 10 0 Time (seconds)

.2 L L)

.8 C

7 . s o

3 0. . . . . .I . . . .

-ie se.ns Pon7eeh:ceo3ln 1nit0( .os l 4 G Loss of AC Power Figure 14.1.11-1 (sheet 1 of 6)

PB,,\IP FSAR (06/0 1" Loss oi All AC Power tO [I-It SLation ALI.Kiliaries Paue 1-1, 1 1 1-ý of ! 5.

PBNP PSAR (06101 Loss o: A)) AC Power 10 Ihe Siation AuuIaries Page )- C)

FiGURE 14. 1.1 -1I ShetL 2 of 6 UNIT I (MIODEL 44F SCG) LOSS OF AC POWER I

0 a

Time. (3ecccnd3)

C L

L 0 L.

Cr 2-I .- C. 0 Time (3econds)

Poinf Seach Nucleoia~n Units I ond 2\

Ur,* I (mo~d. 44.F SC)

Lost of AC Pewor Figure

(--,h*ot Z cri 6)

1000ý 1u00 E

2.)

"=--2000 0

C-,

-5000 10 10 11 Time (seconds) 12 E

0 o ..

0 o .

.4 0

Of CD 0 10 10 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 44F SG)

Loss Gf AC Power Figure 14.1.11-I (sheet 2 of 6)

PBNP FSAR (06/0 1I Los,;. of All AC Power to the Station Au.-61iaric! Page 14. 1.1 1-0 f I FIGURE 14- 1.11-]

Sheet 3 of 6 UNIT I (MIODEL 44F SG'T LOSS OF AC POWER

-- a

-~ - - - - - --- -,

CL Time (second%)

L.

730 0

I 66-0 C,

0 N. -

0 0

C) 5~ - - - - --- -- -- --

e4 54 0 0 1I 5

0 10 Time (secanas)

Point BeachiNucleot- Picini Units t 1and 2 Unit 1 (hWodeI '44F SG)

Less of AC Power Fligu .v ',.

tsI.ef 3 of 6)

I

i 70

" 660 660- ---------- - - - -~

Z 620 CL 0

0 500 I--

1 500 Time (seconds) 06 C- 620 0

08 I 500 0 5W -- -/- - - - -

0 c-J

- 10 10 10 10 10 10 Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 44F SG)

Loss of AC Power Figure 14.1.11-1 (sheet 3 of 6)

PBNP FSAR (06/01) Loss of All AC Power to the Station Auxiliaries Page 14. 1.1 1-7 of 15 PBNP PSAR (06/0]) Los\ of All AC Power to the Station Auxsltarie Page 14.1.1] 7oflS FIGURE 14.1.11-1 Sheet 4 of 6 UNIT I (MODEL 44F SG) LOSS OF AC POWER

-1

\

C.,

0.

I..

tO*~

r \ 10 a., I 00a~

Tjme (second )

I1Ia0C 0 /i

/

L

/

-C laoo .*/

/

I-*

to 3

-l

• 0 CI1 IC.

Ic i Time (seconds)

PQInI Beach Nucle jr Plant Units 1 and 2, Unit 1 (Modal 40F SG) \,

Loss of AC Power Fi(hure 14.1.11

$h let A of 6} "

I

240 0 22Oo-0 1900 1700 0 19000 Time (seconds) 1100 CD I 3 1700 N9 10, 10D 10' 10 10 10 100 7

Time (seconds)

Point Beach Nuclear Plant Units 1 and 2 Unit 1 (Model 44F SC)

Loss of AC Power Rigre 14. 1.11 -1 F.sheet 4 of 6)

P'BN P FS AR 1,06/0 1 Loss of All AC Powecr to the Station Auxiliaries Page 14. 1.11 -8 ofi 15 PBNP PSR ~Ou;Los I of Al AC Poe to the Statio Auiire Iae J4 I-o:

FIGURE 14.111-1 Sheet 5 of 6 I-..

UNIT I (MODEL 44F SG) LOSS OF AC POWER I !

¶ 70 0 Ii r /

I rt1 a.

a V

{ iL

/

1

/' 10 f : ."

• r l ,I I i ' '

0 ' "/: l e '

0 I0 14 T'im e (38copdas) a

!L "o( a- /

'4 aa --

'4 0o I

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Loss of AC Power Figure 14.1.11-1 (sheet 5 of 6)

PBNP FSAR (06/Cl Loss of All AC Power to the Station Auxiliaries Page 14.1 .1 -9 of 15 FIGURE 14.1 AI-1.

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Loss of AC Power Figure (shoot14.1.11-1 6 of 6)

PBNP FSAR (06/01) Loss of All AC Power to the Station Auxiliaries Page- 1A. 1.1 1-10 of 15 PBNP FSAR (06/01) Loss of All AC Power to the Station Auxiliaries 1 III lUoF 5

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Loss of AC Power Figure 14.1.11-2

(,sheet I of 6)

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Loss of AC Power Figure 14.1.11-2 (sheet 2 of 6)

PBNP FSAR (06/01) Loss of All AC Power to the Station Auxiliaries Page !4.1.11-12o1 15 C

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Loss of AC Power Figure 14.1.11-2 (sheet 3 of 6)

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Loss of AC Power Figure 14.1.11-2 (sheet 5 of 6)

PBNP PSAR (06/01) Loss of All AC Powtr to the Station Auxiliaries Page 1411.11-15of 15 C q.p1ll FIGURE 14.1.11-2 Sheet 6 of 6 UNIT 2-(IJELTA-.47 SG) LOSS OF AC POWER N

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Point Beach Nuclear Plant Units 1 and 2 Unit 2 (Delta-47 SG)

Loss of AC Power Figure 14.1.11-2 (sheet 6 of 6)