3F0602-05, Response to Request for Additional Information 263, Revision 0, Relocation of Reactor Coolant System Parameters to Core Operating Limits Report & 20 Percent Steam Generator Tube Plugging

Response to Request for Additional Information #263, Revision 0, Relocation of Reactor Coolant System Parameters to Core Operating Limits Report & 20 Percent Steam Generator Tube Plugging

ML021690261
Person / Time
Site:	Crystal River
Issue date:	06/05/2002
From:	Roderick D Florida Power Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0602-05 32-1257514-01, 51-5000475-01
Download: ML021690261 (58)

Text

j Florida Power AProgress Energy Company Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Ref: 10CFR50.90 June 5, 2002 3F0602-06 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3, Response to Request for Additional Information LAR

1. 263, Revision 0, Relocation of Reactor Coolant System Parameters to the Core Operating Limits Report and 20 percent Steam Generator Tube Plugging

References:

1. NRC to FPC letter, 3N1201-10, dated December 27, 2001, "Crystal River Unit 3 Request For Additional Information RE: Proposed License Amendment Request No.263, Revision 0, Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging" (TAC NO. MB2499)

2. NRC to FPC letter, 3N0402-19, dated April 23, 2002, "Crystal River Unit 3 Request For Additional Information RE: Proposed License Amendment Request No.263, Revision 0, Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging" (TAC NO. MB2499)

3. FPC to NRC letter 3F0701-11, dated July 24, 2001, "License Amendment Request #263, Revision 0, Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging"

Dear Sir:

Florida Power Corporation (FPC) submits the additional information requested in References 1 and 2 concerning the License Amendment Request #263, Revision 0 (Reference 3).

Attachment A provides the responses to the Request for Additional Information (RAI).

Attachment B and Attachment D provide Framatome ANP calculations on flow induced vibration for steam generator tubes. These calculations provide additional information concerning the RAI in Reference 2.

Attachment D contains information that Framatome ANP considers to be proprietary.

Framatome ANP requests that the proprietary information in this response (Attachment D) be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4), 2.790(a)(4) and 2.790(d)(1). An affidavit supporting this request is provided in Attachment C. ý,10 15760 West Power Line Street

• Crystal River, Florida 34428-6708 * (352) 795-6486

U.S. Nuclear Regulatory Commission 3F0602-06 Page 2 of 3 This letter makes no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Sid Powell, Supervisor, Licensing and Regulatory Programs at (352) 563-4883.

• Daniel L. Roderick Director Site Operations Crystal River Nuclear Plant DLR/pei Attachments:

A. Response to NRC Request for Additional Information B. FRA-ANP Engineering Information Record, CR-3 OTSG FIV Margins, document number 51-5000475-01, October 1, 2001, Non-Proprietary C. Framatome ANP Affidavit of Proprietary Information D. FRA-ANP Calculation file, "Flow Induced Vibration Analysis of TMI OTSG Tubes Due to Power Uprate," document number 32-1257514-01, June 23, 1997, PROPRIETARY xc: Regional Administrator, Region II w/o Attachment D Senior Resident Inspector w/o Attachment D NRR Project Manager

U.S. Nuclear Regulatory Commission 3F0602-06 Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS Daniel L. Roderick states that he is the Director Site Operations, Crystal River Nuclear Plant for Progress Energy; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief. Z /7 Daniel L. Roderick Director Site Operations Crystal River Nuclear Plant The foregoing document was acknowledged before me this 54-h day of

,*-hr , 2002, by Daniel L. Roderick.

792 Signature of Notary Public State of Fkrg*da 4 Susan I. MicDonald Expkr= 14 200

&ofth Ga. Iva (Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known -OR- Identification V

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ATTACHMENT A LICENSE AMENDMENT REQUEST #263, REVISION 0 Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging Response to NRC Request for Additional Information

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 1 of 29 Response To NRC Request For Additional Information

1. NRC Request:

The license amendment request is to allow plugging of up to an equivalent of 20 percent of all Once-Through Steam Generator (OTSG) tubes. For symmetric plugging, this allows up to 20 percent of the tubes in each OTSG to be plugged.

Florida Power Corporation did not specifically state the level of asymmetric tube plugging or the maximum percentage of plugged tubes in anyone OTSG that is to be allowed. Although asymmetric tube plugging analyses and results were discussed, it is not clear which analysis determines the limiting acceptable tube plugging asymmetry.

Please provide a discussion regarding any restrictions being placed on the maximum number of plugged tubes in any one OTSG and the maximum plugging asymmetry between OTSGs. Provide justification for these restrictions.

FPC Response:

Relative to OTSG tube plugging, the license amendment request will allow up to an equivalent of 20 percent of the tubes in either or both steam generators to be plugged, including 0 percent/20 percent asymmetric tube plugging. Framatome ANP performed the analysis in such a manner as to eliminate the need for an asymmetric plugging limit. There is no single analysis that defines the maximum allowed asymmetry. Steam generator tube plugging reduces the tube heat transfer area and decreases the flow area through the tubes.

These changes alter the fluid temperatures, reactor coolant system (RCS) flow rates, and secondary side mass inventories during steady-state operation and transients. Asymmetric tube plugging effects create both overall and loop specific changes that are scenario specific. The overall RCS steady-state changes are best represented by an equivalent overall plugging limit (i.e., change in total core flow), but these may have asymmetric effects reflected by loop and core inlet temperature differences that may create slight core power tilts. The transient analyses in which the loops remain roughly symmetrical (e.g.,

four-pump coastdown, turbine trip, and loss of all main feedwater) are also best characterized with symmetrical plugging. Transients for which asymmetric effects are important (e.g., cold leg loss-of-coolant accidents (LOCAs), locked pump rotor, single pump coastdown) are best characterized with asymmetric plugging. For these events, specific analyses were performed to address the asymmetric effects. For the LOCAs, the peak clad temperature calculations considered up to 25 percent plugged tubes in one steam generator with up to 15 percent in the opposite loop steam generator (this distribution was most limiting for the LOCA analysis). For the departure from nucleate boiling (DNB) and core peaking aspects, the core flow distribution created by a 0 percent and 30 percent tube plugging asymmetry was evaluated. It was determined that this asymmetry could reduce the localized core inlet flow velocity by 0.4 percent. This asymmetric effect on flow was included in the global flow reduction associated with 20 percent tube plugging in both OTSGs. (A further discussion of the asymmetrical effects on the DNB analysis is

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 2 of 29 presented in the response to Question 2.e. below.) Based on these analyses, no asymmetric plugging limits are required. The OTSG plugging limit is a maximum equivalent of 20 percent of the tubes that can be plugged in either or both steam generators.

2. NRC Request:

Three vendor-proprietary OTSG Thermal-Hydraulic performance-related computer codes were referenced and used by the licensee in the 20 percent tube plugging analyses:

FSPLIT VAGEN PORTHOS

a. Discuss the interaction of these three codes with the design-basis accident/transient analysis codes.

FPC Response:

There is no direct interaction between these three codes and the design basis accident/transient analysis codes. Indirect interaction has included comparing the results from these three design codes with the results from the design basis accident/transient analysis codes.

b. These three codes have not been reviewed and approved for use by the U.S.

Nuclear Regulatory Commission. Are these codes appropriately controlled in accordance with Title 10 of the Code of Federal Regulations, Appendix B?

FPC Response:

Yes, these codes are used in accordance with Framatome ANP procedures governing the use of internally developed design codes. There is no requirement for NRC review and approval of these types of codes.

c. Confirm that the input data, model, and options selected accurately represent CR-3 plant-specific design features and justify the use of these codes for CR-3.

Please note that the use of the PORTHOS code was not discussed in the submittal.

FPC Response:

The FSPLIT model (B&W 32-1203121-01, "FSPLIT Certification Analysis") is a Crystal River Unit 3 (CR-3) specific hydraulics model that has been benchmarked to measurements derived from RCS flows and incorporates CR-3 specific reactor coolant pump (RCP) performance curves and CR-3 hydraulics inputs. The VAGEN model (FTI Report TRG 78-2, Revision K, "VAGEN Generator Performance Prediction,"

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 3 of 29 May 1995) is generic to B&W design 177 fuel assembly (FA) plants. It has been benchmarked to model boiler and plant data. The PORTHOS model (FTI 32-1239294-00, "PORTHOS for OTSG Applications") is generic to non-internal auxiliary feedwater (AFW) header, B&W 177 FA design plants, and thus applies to CR-3. PORTHOS has been benchmarked to model boiler and plant data.

d. Provide a discussion of the key assumptions and methodology used to determine the impact of symmetric and asynmnetric OTSG tube plugging on RCS flow and hot-to-cold leg temperature difference (delta-T) for both three and four reactor coolant pump (RCP) operation and confirm that the results are bounding.

FPC Response:

The methodology used in calculating the RCS flow changes due to tube plugging was to reduce the steam generator flow area and adjust the tube sheet inlet and outlet form loss factors commensurate with the degree of tube plugging. Both symmetric and asymmetric plugging were modeled within the RCS hydraulics model to define the reactor vessel inlet nozzle flow rates. Experimental data from vessel model flow tests were used to correlate inlet nozzle flow asymmetry (due to OTSG tube plugging asymmetry) with core inlet velocity asymmetry. The results of the analyses showed that asymmetric plugging resulted in applying an additional 0.5% decrease in core inlet flow to compensate for its effects.

There is no limit on the hot-to-cold leg temperature difference. The integrated control system (ICS) will re-ratio feedwater flows to the OTSGs to ensure equal cold leg temperatures.

e. Provide a discussion regarding the methodology used to calculate, combine, and apply the symmetric and asymmetric core design and Departure from Nucleate Boiling (DNB) flow penalties.

FPC Response:

The impact of OTSG tube plugging on the RCS flow rate was quantified for the CR-3 plant as well as other B&W-designed 177 FA plants as part of a B&W Owners Group project to provide supporting justification for 20% tube plugging. The analyses indicated that CR-3 would have a global RCS flow rate reduction of 4.0% of design flow by going from a 0% tube plugging condition to a 20% tube plugging condition (as seen in Table 1, Section C.1 of Reference 2, Attachment F). The maximum global RCS flow rate reduction for all 177 FA plants examined was 4.5%. Additional analyses found that a bounding asymmetric OTSG plugging condition (30% in one OTSG and 0% in the other OTSG, or 30%/0%) could reduce a localized core inlet flow velocity by 0.4%. However, the same bounding asymmetric plugging condition only has a 3.5% impact on the global RCS flow reduction. Since the maximum global RCS

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 4 of 29 flow rate reduction (associated with 20%/20%) can not occur at the same time as the most severe localized core inlet flow velocity reduction (associated with a bounding asymmetric tube plugging condition) it was concluded that a 4.5% global RCS flow rate reduction impact for DNB analyses conservatively bounds the RCS global and localized core inlet flow effects for all the 177 FA plants examined.

In the CR-3 DNB analyses supporting 20% tube plugging, the 4.5% global RCS flow rate reduction has been completely incorporated. This resulted in the reduction of the nominal RCS flow rate basis, for DNB analyses, from 109% of design flow rate to 104.5% of design flow rate.

3. NRC Request:

Regarding the impact of tube plugging on increased RCS flow resistance, please provide a reference and discuss the method used to quantify that 6.7 sleeves have the same hydraulic resistance as one plug. Include in the discussion any test data and analyses this is based on, the effects of OTSG tube plugging asymmetry on this value, and whether this equivalent bounds all steady-state and transient operating conditions.

FPC Response:

The 6.7 sleeve-to-plug ratio is based on the most prevalent sleeve design. This value was provided as an example only and was not intended to imply that this value would be used for all sleeves at CR-3. Other sleeve designs would have different sleeve-to-plug ratios.

The sleeve-to-plug ratio is based on adjusting the steam generator hydraulic resistance (within the CR-3 specific FSPLIT model) based on actual sleeve designs used at CR-3.

This includes a reduction in tube flow area and decrease in frictional diameter over the sleeve length and accounting for the additional contraction into the sleeve and expansion out of the sleeve. The RCS flow changes due to sleeving was equated with the RCS flow changes due to plugging. This methodology is currently being used for CR-3 and will not change as a result of the proposed amendment.

4. Regarding the impact of reduced RCS flow rate and potential asymmetry in system flow on fuel component integrity:

a. Clad oxidation has been determined to be the limiting constraint while CR-3 utilizes Zircaloy-4 cladding. Discuss the administrative controls, processes, or procedures in place to ensure the necessary limitations on burnup and peaking for cycle-specific reload designs are such that acceptance criteria for clad oxidation will be maintained.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 5 of 29 FPC Response:

Cycle specific checks of clad oxidation levels are performed as part of Framatome's standard reload licensing analysis activities following methods outlined in the NRC approved topical report BAW-10186P-A Rev. 1, "Extended Burnup Evaluation."

These cycle specific checks are an integral part of the cycle design process in which the highest burnup fuel pins in each fuel batch are analyzed using Framatome's KOROS/COROS2 evaluation model. The cycle specific check incorporates bounding core parameters, while typically employing a pin specific power history. By performing these checks early in the reload licensing process, the validity of the proposed cycle design is confirmed.

b. Please discuss the expected increase in clad oxidation levels for the current CR-3 Zircaloy-4 fuel type and demonstrate acceptance criteria are met.

FPC Response:

CR-3 Cycle 13 is the first reload cycle to consider the effects of 20% OTSG tube plugging. Results from the Cycle 13 clad oxidation check have shown that, although the highest burnup pins are approaching the clad oxide limit, acceptable results have been demonstrated by the use of pin specific power histories as described in BAW-10186P-A Rev. 1. Cycle design strategies for future cycles are expected to be similar to those of Cycle 13 resulting in similar peak burnups and similar pin power histories. It is, therefore, expected that future detailed cycle specific checks using pin specific power history data will continue to yield acceptable clad oxidation results.

c. Discuss the administrative controls, processes, or procedures in place to ensure the necessary limitations on clad lift-off for cycle-specific reload designs are such that acceptance criteria will be maintained.

FPC Response:

Cycle specific checks of fuel thermal performance, including inception of pellet/clad lift-off, are performed as part of Framatome's standard reload licensing analysis activities following methods outlined in the NRC approved topical reports BAW-10162P-A, "TACO3 Fuel Pin Thermal Analysis Computer Code,"

BAW-10184P-A, "GDTACO - Urania Gadolinia Fuel Pin Thermal Analysis Code,"

and BAW-10183P-A, "Fuel Rod Gas Pressure Criterion." These cycle specific checks confirm that the peak burnup pin in each fuel batch is bounded by a generic fuel performance analysis. In those instances where bounding generic analyses are too restrictive, additional margin is obtained by evaluating fuel performance using pin specific power history data. Cycle 13 was the first reload cycle to consider the effects of 20% OTSG tube plugging at CR-3. For Cycle 13, the reduced flow rate associated with the higher tube plugging was incorporated into all fuel performance analyses and all acceptance criteria were met.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 6 of 29

d. Discuss the methodology applied for the guide tube boiling analysis to determine that no saturation will occur in the guide tube Assembly Hold-down Springs, Guide Tubes, and Spacer Grids.

FPC Response:

The guide tube boiling analysis is performed using a thermal-hydraulic methodology which conservatively models a limiting control component within the guide tube and determines the margin to saturation. This analysis is a deterministic type analysis where key parameters such as total core power, flow rate within the guide tube, and heat generation within the guide tube are treated at a conservative level. The generic analysis, which used design assumptions that were bounding for all B&W 177 FA type plants, conservatively bounded the CR-3 specific design condition. By demonstrating positive margin to saturation for this conservative set of design conditions, assurance is given that CR-3 operation with up to 20% tube plugging will be acceptable.

Additional fuel mechanical evaluations were supported by LYNXT based calculations which determined that the impact of tube plugging was a less than 2°F increase in coolant temperatures at the spacer grid locations and at the assembly exit (e.g. at the assembly hold down springs). These LYNXT calculations were also based on design assumptions which were bounding for all B&W 177 FA type plants (BAW-10156-A, Rev.1, "LYNXT - Core Transient Thermal Hydraulic Program," March 1993). The fuel mechanical effects of this increase in temperature at the spacer grids and hold down springs were evaluated and shown to be bounded by the effects that would be seen within the guide tubes.

5. NRC Request:

Regarding the Core Physics impacts:

a. Discuss the rationale for choosing a +/-1.1 0F inlet temperature asymmetry penalty.

Is this meant to be consistent with the requested TS change of 1.2 OF?

FPC Response:

The licensing amendment submittal requested a change of 1.2°F to the RCS hot leg temperature DNB surveillance parameter of SR 3.4.1.2. There is no direct connection between that parameter and the +/- 1. I°F inlet temperature asymmetry value. The +/- 1. I°F inlet temperature asymmetry was associated with evaluation of potential core power distributions that could be produced by asymmetric steam generator tube plugging. The full-core NEMO power distribution analysis actually used a value of 2.2°F and split it so that the core inlet temperature from one side was higher than nominal by 1. 1°F and the other side was lower than nominal by 1. 10F. Since a +/- 1. 1IF delta Tin (for at total

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 7 of 29 of 2.2 0 F loop delta-T) due to asymmetric plugging was modeled, the calculations remain bounding for the worst case (maximum) of 1.7°F loop delta-T. The latter figure actually represented a case that was more severe than 20%/0% tube plugging.

Furthermore, the power distribution evaluation did not take credit for any ICS action to mitigate the loop delta-T. The ICS would normally drive the loop delta-T to zero (or the desired setpoint) even with asymmetric plugging. Therefore, in reality, it was concluded that there would be no delta-T penalty for asymmetric plugging, thus using the +/- 1. IVF value is conservative.

b. Provide more detail regarding the "conservative equation" to determine an adjusted burnup which was developed for use in core designs.

FPC Response:

Some of the power peaking limits are burnup-dependent (examples: LOCA limits and linear heat rate limits based on transient cladding strain criteria). Therefore, an investigation was performed to determine if the differences in fuel rod burnups that could accumulate with asymmetric steam generator tube plugging could be significant.

It was found that the maximum cumulative burnup increases and decreases were on the order of 50 megawatt days per metric ton of uranium (MWd/mtU) (approximately 1.5 Effective Full Power Days) and are not large enough to cause concern with regard to application of burnup-dependent peaking limits in licensing analyses. Consequently, it was concluded that simulation of the core power distribution using quarter-core symmetry could be continued in reload cycle licensing evaluations.

c. Please provide a listing of the approved methods used to evaluate the effects of OTSG tube plugging on each aspect of core physics discussed in the submittal.

FPC Response:

The core physics analyses were performed in accordance with Section 5 of BAW 10179P-A, which is Framatome ANP's approved safety criteria and methodology topical report for reload safety evaluations. The power distribution simulations were performed with Framatome ANP's standard nuclear design code (NEMO) used for cycle-specific reload safety evaluations. The NEMO code is fully described in an approved topical report, and has been subjected to an extensive verification program that quantified the uncertainties associated with its application (BAW-10180-A, Rev. 1).

The code was used within its range of applicability to simulate core power distributions used in support of the evaluation of OTSG tube plugging.

e. Provide a quantitative discussion regarding the impact on transient xenon power distribution.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 8 of 29 FPC Response:

Transient xenon power distributions are simulated in order to evaluate skewed axial power distributions in reload safety evaluations. Section 5 of BAW-10179P-A describes how the transients are simulated and used to generate limiting core power distributions. The evaluations performed for steam generator tube plugging simulated the same types of transients that are used in cycle-specific reload safety evaluations for B&W units, however full-core simulations were generated in order to account for power distributions with asymmetric tube plugging. The correlations of power peaking to axial offset were found to remain similar to those generated for symmetric cores with only small changes noted in axial power offset and peaking factors.

e. Do the four configurations evaluated for each core physics aspect bound the maximum allowed plugging asymmetry (this ties in with question 1)?

FPC Response:

The four configurations were designed to provide a bounding evaluation of core power distributions and margins to power peaking limits for both symmetric and asymmetric power distributions with 20% steam generator tube plugging. The base case (no tube plugging) was included in order to generate data for comparative analysis.

6. NRC Request:

Please provide the following information for the four RCP and one RCP Coastdown, Locked Rotor, Startup Event, Loss of All AC Power/Station Blackout (SBO),

Anticipated Transient Without Scram (ATWS), Large and Small Break Loss of Coolant Accidents and Main Steam Line Break events, which were either re-analyzed or re-evaluated for OTSG tube plugging:

a. Initial conditions and justification including instrument uncertainty discussions.

b. Sequence of events.

c. Trip setpoints including instrument uncertainty discussions.

d. Safety system actuation setpoints including instrument uncertainty discussions.

e. Single failure criterion.

FPC Response for A through E:

Each of the events are addressed separately below:

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 9 of 29 Four Reactor Coolant Pump (RCP) Coastdown No new system analysis was required for the four-pump coastdown accident to justify increased levels of OTSG tube plugging. Only the DNB portion of the event was reanalyzed. The analysis method is outlined in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," Revision 2. The original system response provides input to the DNB analysis that includes the transient core exit pressure, core inlet temperature, core inlet flow rate, and the total core power values. The system analysis was performed at the minimum design flow rate versus the minimum DNB flow rate to ensure a conservative pressure and temperature prediction. The core inlet flow is normalized and, therefore, will not be affected by the increased levels of OTSG tube plugging. The core power response is also normalized since the initial core power level in the calculation accounts for the heat balance uncertainty. Based on the results of the system response calculation, the core pressure and the core inlet temperature are held constant throughout the DNB calculation at the initial nominal values. The uncertainty in these parameters is included in DNB calculation through application of the statistical core design methodology (BAW-10187P-A, "Statistical Core Design for B&W-Designed Plants"). The minimum DNB flow rate is reduced to account for 20 percent OTSG tube plugging and is applied to the normalized flow data from the system analysis. The loss of flow is the initiating event and the reactor is tripped based on the loss of all pumps reactor protection system (RPS) trip with a 2.6-second time delay. Consistent with the plant design basis, there are no single failures imposed on this analysis. No safety system actuation or operator actions are required to mitigate this event.

The transient normalized flow and power response is provided in Figure 1. The DNB response is provided in Figure 2 and was generated using the LYNXT computer code (BAW-10156A, Rev. 1, "LYNXT - Core Thermal-Hydraulic Program, Revision 1"). For this event, the calculated DNB Ratio (DNBR) remains above the thermal design limit for all times. The current statistically based thermal design limit for CR-3 is 1.40 BWC. All of the restrictions and limitations of the approved computer codes and methods were met for the four-pump coastdown analysis.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 10 of 29 Figure I Figure 2 CR-3 Four Pump Coastdown -20% OTSG Tube Plugging CR-3 Four Pump Coaddown -20%OTSG Tube Plugging 1.2- 220 1 0?_0 0.8-* ZOO

• ~0.6 1.9 0 z 1.90 0.2  ; Care I. 1.75e I- RCS Flaw 1.70________________

3.5 4 0.0 0.5 1.0 1.5 2.0 25 3.0 3.5 4.0 0 0-5 1 1.5 2 2.5 3 "Time (sec) Time (sec)

One Reactnr Cnnolnt rnmp (RCP) COaqtdown No new system analysis was required for the single-pump coastdown accident to justify increased levels of OTSG tube plugging. Only the DNB portion of the event was reanalyzed. The analysis method is outlined in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," Revision 2.

The original system response provides input to the DNB analysis that includes the transient core exit pressure, core inlet temperature, core inlet flow rate, and the total core power values. The system analysis was performed at the minimum design flow rate versus the minimum DNB flow rate to ensure a conservative pressure and temperature prediction.

The core inlet flow is normalized and therefore will not be affected by the increased levels of OTSG tube plugging. The core power response is also normalized since the initial core power level in the calculation accounts for the heat balance uncertainty. Based on the results of the system response calculation, the core pressure and the core inlet temperature are held constant throughout the DNB calculation at the initial nominal values. The uncertainty in these parameters is included in DNB calculation through application of the statistical core design methodology (BAW-10187P-A, "Statistical Core Design for B&W Designed Plants"). The minimum DNB flow rate is reduced to account for 20 percent OTSG tube plugging and is applied to the normalized flow data from the system analysis.

The loss of flow is the initiating event and the reactor is tripped based on the flux-to-flow reactor protection system (RPS) trip setpoint of 1.12985 with a 2.18-second time delay.

Consistent with the plant design basis, there are no single failures imposed on this analysis.

No safety system actuation or operator actions are required to mitigate this event.

The transient normalized flow and power response is provided in Figure 3. The DNB response is provided in Figure 4 and was generated using the LYNXT computer code (BAW-10156A, Rev. 1, "LYNXT - Core Thermal-Hydraulic Program, Revision 1"). For this event, the calculated DNBR remains above the thermal design limit for all times. The

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 11 of 29 current statistically based thermal design limit for CR-3 is 1.40 BWC. All of the restrictions and limitations of the approved computer codes and methods were met for the single-pump coastdown analysis.

Figure 3 CR-3 Single Pump Coasbdown - 20%OTSG Tube Plugging Z

0 1 2 3 4 5 6 7 8 Time (sac)

I_oclced Thimp Rotor A locked rotor event is the reduction in forced flow through the reactor coolant system when the rotor of one of the reactor coolant pumps seizes. When the rotor seizes, forced flow is no longer provided by the affected pump, which results in a more rapid initial reduction in flow than the four-pump coastdown event. The reactor is tripped based on the flux-to-flow trip function of the reactor protection system. Once the control rods begin to insert, the core power is reduced and the minimum DNBR begins to increase. Consistent with the CR-3 design basis, there are no single failures assumed in the analysis. There is no safety system actuation or operator actions required to mitigate this event. A list of the key parameters that is used in the analysis is provided in the following table.

Parameter Value Core Power 1.02 x 1.001 of 2568 MWt Primary Side Ave Fluid Temp. 579 OF RCS Pressure @ HL Tap 2170 psia Moderator Coefficient at Full Power +0.0 E-4 (Ak/k)/° F Doppler Coefficient at Full Power -1.17 E-5 (Ak/k)/°F Reactor Trip on Flux-to-Flow Ratio 1.12985 %FP/ %Flow Reactor Tip Time Delay 2.18 sec

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 12 of 29 In order to support increased levels of OTSG tube plugging, the system response to the locked pump rotor event was reanalyzed using the RELAP5/MOD2-B&W computer code (BAW-10164P-A, "RELAP5/MOD2-B&W, An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis," Revision 3).

RELAP5/MOD2-B&W has been approved by the NRC for non-LOCA applications (BAW-10193P-A, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors"). All of the restrictions and limitations of the approved topical reports were met. Once the system calculation was completed, the DNB portion of the event was reanalyzed. The DNB analysis method is outlined in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," Revision 2.

The system response calculation provides input to the DNB analysis that includes the transient core exit pressure, core inlet temperature, core inlet flow rate, and the total core power values. The system analysis was performed at the minimum design flow rate versus the minimum DNB flow rate to ensure a conservative pressure and temperature prediction.

The core inlet flow is normalized and is provided in Figure 5. The core power response is also normalized since the initial core power level in the calculation accounts for the heat balance uncertainty and is shown in Figure 5. Based on the results of the system response calculation, the core pressure and the core inlet temperature are held constant throughout the DNB calculation at the initial nominal values. The uncertainty in these parameters is included in the DNB calculation through application of the statistical core design methodology (BAW-10187P-A, "Statistical Core Design for B&W-Designed Plants").

The minimum DNB flow rate is reduced to account for 20 percent OTSG tube plugging and is applied to the normalized flow data from the system analysis. The DNB response is provided in Figure 6 and was generated using the LYNXT computer code (BAW-10156A, Rev. 1, "LYNXT - Core Thermal-Hydraulic Program, Revision 1"). For this event, the calculated DNBR remains above the thermal design limit for all times. The current statistically based thermal design limit for CR-3 is 1.40 BWC. All of the restrictions and limitations of the approved computer codes and methods were met for the locked rotor analysis.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 13 of 29 Figure 5 CR4 Locked RCP Rotor - 20% OTSG Tube Plugging 1.2000 1.0000 0.8000 0.6000 0

0.4000 0.2000 0.0000 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 Time (sec)

Stamrtu vent The startup event is a rod withdrawal accident from subcritical or low power conditions.

The rapid reactivity addition causes the RCS pressure, temperature, and core power to increase. The control rods are inserted on a reactor trip based on either high system pressure or high flux. Once the control rods begin to insert, the event is quickly terminated. A list of key parameters used in the analyses is provided in the following table.

Startup Event Key Parameters Initial Conditions Core Power 2.568 W Hot Leg Pressure 2170 psia RCS Average Temperature 532 0F Reactor Protection System 112% of 2,568 MWt?()

High Flux Reactor Trip Setpoint 2400 psia (Hot Leg) ')

High RCS Pressure Reactor Trip Setpoint 0.4 sec ()

Delay Time for High Flux Reactor Trip 0.6 sec ()

Delay Time for High RCS Pressure Trip

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 14 of 29 Startup Event Key Parameters Control Rod Drive Assembly (CRA)

Maximum Control Rod Speed 30 in/min Maximum Number of CRAs 60(a)

Maximum Rod Worth, all rods 12.9% Ak/k(a)

Maximum Reactivity Addition rate, 9.27 E-4 (Ak/k)/sec all rods at maximum speed Nominal Rod Worth of single group, 3.0% Ak/k when reactor is critical Nominal Reactivity Addition Rate, single rod group 2.15 E-4 (Ak/k)/sec Doppler Coefficient at Rated Power -1.3 E-5 (Ak/k)/OF (b)

Moderator Coefficient at Zero Power +0.9 E-4 (Ak/k)/!Fc(b)

Control Rod Travel Time, to 2/3 Insertion 1.4 sec Minimum Tripped Rod Worth Used 2.43 % Ak/k(c)

NOTES (a) The number of CRAs is/was not explicitly modeled in the analysis. The input to the analysis is a reactivity insertion rate that is independent of the number or worth of the control rods. For each new core reload, the reactivity insertion rate is converted to a control rod worth and is compared to the cycle specific value to verify that the safety analysis remains bounding.

(b) In order to support increased OTSG tube plugging limits, a new analysis was performed with 30 percent of the OTSG tubes plugged (bounds 20 percent) using RELAP5/MOD2-B&W.

(c) Sufficient control rod worth was added to provide 1% (Ak/k) shutdown margin at hot zero power conditions.

In order to support increased levels of OTSG tube plugging, a spectrum of reactivity insertion rates were reanalyzed using the RELAP5/MOD2-B&W computer code (BAW 10164P-A, "RELAP5/MOD2-B&W, An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis," Revision 3). RELAP5/MOD2-B&W has been approved by the NRC for non-LOCA applications (BAW-10193P-A, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 15 of 29 Reactors"). All of the restrictions and limitations of the approved topical reports were met. The reanalysis supports up to 30 percent (equivalent) steam generator tube plugging and a pressurizer safety valve lift tolerance of +/-3 percent. The reanalysis was performed using the most positive moderator temperature reactivity coefficient at hot zero power conditions and a least negative Doppler reactivity coefficient as shown in the table above.

Consistent with the methods described in Section 14.1.2.2. of the CR-3 Final Safety Analysis Report (FSAR), the reactivity insertion rate was varied to find the limiting value with respect to peak system pressure and peak power. Also consistent with the CR-3 design basis, no single failures are modeled in the analysis. For the limiting reactivity insertion rate, the peak reactor coolant pressure was shown to be less than 110 percent of the design pressure (or < 2750 psig) using a pressurizer safety valve lift setpoint of 2500 psig with a +3 percent lift tolerance. The peak power was less than 112 percent of 2568 MegaWatts thermal (MWt). Therefore, all the acceptance criteria were met. Based on these results, it is concluded that the reactor is completely protected against any startup accident involving the withdrawal of any or all control rods and with 3 or 4 reactor coolant pumps in operation. In no case does the thermal power exceed the design overpower condition, and the peak pressure never exceeds code allowable limits.

The pressure response for the startup event was provided on Page 20 of Attachment F to the License Amendment Request No. 263. The peak power response for the spectrum of reactivity insertion rates analyzed is provided in Figure 7.

Figure 7 CR-3 Startup Event- 20%SG Tube Plugging Thermal Power) 80 -(Peak 70 -1 60 S50 sO S40 S30 20 a.

10 0

0.75 1.00 1.25 1.50 1.75 2.00 2.25 Reactivity Insertion Rate (x10-4 dkl/lsec)

I osp of All AC Power The loss of all AC power event results in a reactor trip and loss of all reactor coolant pumps immediately upon loss of power. The transient is initiated from nominal conditions.

The net energy addition to the primary coolant during the transient is less than that during a loss of main feedwater accident. Thus, the consequences of the loss of all AC power

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 16 of 29 accident would also be bounded by the loss of main feedwater accident. The only safety actuation function is to initiate emergency feedwater (EFW) on low OTSG level. Since EFW must absorb less energy because the reactor coolant pumps are not operating and because the minimum EFW flow requirement was not challenged due to increasing the level of OTSG tube plugging in the loss of main feedwater accident, no new analysis was required.

Station Blackout (SBO)

A generic B&W-plant analysis was originally performed to address the station blackout issue in accordance with NUMARC 87-00. This transient was not part of the original design basis of the B&W-designed plants, but was included in the CR-3 FSAR for information. Nonetheless, as part of a B&W Owners Group project, FRA ANP Document 51-5009660-01, "Evaluation of Extended Tube Plugging Limits for the Once-Through Steam Generator," the effect of increased tube plugging on the generic calculation was performed. The generic analysis was based on a conservative power level of 2772 MWt verses the CR-3 rated power level of 2544 MWt. Nominal flow, pressure, and temperature conditions were also modeled. A summary of the analyses and bases verifications done per NUMARC 87-00 is documented in letters to the NRC dated April 17, 1989 and March 30, 1990. The NRC's response is documented in the Safety Evaluation Report, dated August 23, 1990.

In addition to the conservative power level that was used in the analysis, an evaluation of increased OTSG tube plugging was performed. The original analysis of this event (assuming 100 gpm total RCP seal leakage plus 11 gpm identified and unidentified leakage) shows that loop flow interruption occurs after the pressurizer empties and the u-bend regions of the hot legs void. Since the liquid volumes of these regions are unaffected by OTSG tube plugging, the time that the loop flow is interrupted will also not change with increased levels of tube plugging. Following loop flow interruption, the loss in inventory through the RCP seal leaks results in voiding of the reactor vessel down to the hot leg elevation and subsequent voiding of the cold leg discharge piping as steam produced in the core passes through the reactor vessel vent valves and depresses reactor vessel downcomer water level. Voiding of the cold leg discharge piping allows the RCP seal leakage to become two-phase, significantly reducing the rate of mass loss from the system. As the reactor vessel head region and cold leg discharge piping are voided, hot-leg and steam generator levels continue to decrease, and the level of the reactor coolant in the steam generators falls below the tube sheet. This allows steam in the primary system to condense on tubes cooled by emergency feedwater flow. This condensation provides removal of core decay heat from the primary system and allows the primary system pressure to decrease. Shortly after the onset of this boiler-condenser mode (BCM) of cooling, the required coping duration (four hours) is met, with the core still covered. Because the liquid volumes of the reactor vessel, cold leg piping, hot leg piping and OTSG upper plenna are unaffected by SG tube plugging, the time to reduce primary water level below

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 17 of 29 the upper SG tube sheets is unaffected by increased levels of tube plugging. The onset of BCM is then unaffected. Although the OTSG heat transfer area for condensation is reduced with tube plugging, as demonstrated in the loss of main feedwater accident, there will still be sufficient tube surface area to boil the emergency feedwater. Consequently, the steam condensation rate via boiler-condenser cooling is sufficient to keep the core protected throughout the coping time. Therefore, OTSG tube plugging will have no effect on the SBO transient. The original analysis remains valid and no new analysis is required.

Anticipated Transient Without Scram (ATWS)

A generic B&W-plant analysis was originally performed to address the ATWS issue. This transient was not part of the original design basis of the B&W-designed plants, and is not included in the CR-3 FSAR. Nonetheless, as part of a B&W Owners Group project, FRA ANP Document 51-5009660-01, "Evaluation of Extended Tube Plugging Limits for the Once-Through Steam Generator," the effect of increased tube plugging on the generic calculations was performed. For the B&W plant design, the loss of feedwater represented the most severe ATWS transient. The original ATWS analysis was performed with no reactor trip, nominal plant conditions, and no single failures. The peak pressure for CR-3 was 3943 psia. The predicted RCS pressure was deemed by the NRC to be too high. As a result, all of the B&W-designed plants were required to install a diverse scram system (DSS) and the ATWS Mitigation System Actuation circuitry (AMSAC) to meet the requirements of Title 10 of the Code of Federal Regulations, Part 50.62. These systems provide independent reactor trip, turbine trip, and EFW actuation signals. In the evaluation for extended tube plugging limits, it was assumed that the DSS and AMSAC functions were available. Consequently, the ATWS transient would become similar to the loss of main feedwater accident, except for a delayed reactor trip. Similar to the loss of main feedwater accident, it was concluded that there is sufficient tube surface area, with OTSG tube plugging, to boil the EFW completely. Since OTSG tube plugging does not have an affect on the minimum EFW flow requirement, adequate heat removal from the RCS would be provided. The peak pressure will be limited by the pressurizer safety valves. Therefore, the RCS peak pressure will meet the ASME Service Level C limit with or without OTSG tube plugging included. As a result, no new analyses were performed for ATWS and the original licensing submittal would remain valid for the analysis of record for CR-3 with or without consideration of OTSG tube plugging.

Large and Small Break Loss of Coolant Accidents The LOCA analyses were performed in accordance with the NRC-approved BWNT LOCA Evaluation Model (BAW-10192P-A, "BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants"). All of the restrictions and limitations of the approved methods were met for these analyses. The LOCA analyses results are summarized in Section 14.2.2.5 of the CR-3 FSAR. The results for the limiting small break LOCA (SBLOCA) transient are provided in FSAR Tables 14-67, 14-68, and 14-69

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 18 of 29 and Figures 14-68, 14-69, and 14-70. The large break LOCA (LBLOCA) results were also reported to the NRC in letter 3F1199-01, dated November 10, 1999, "Notification of Change in Peak Clad Temperature for Small Break Loss of Coolant Accident in Accordance with 10 CFR 50.46(a)(3) and Change in the Analysis of Record for Large Break Loss of Coolant Accident." The analyses considered asymmetric OTSG tube plugging of 25 percent in the broken loop OTSG and 15 percent in the intact OTSG (bounds 20%/20% and 20%/0%). The calculations demonstrated that all of the LOCA-related acceptance criteria contained in Title 10 of the Code of Federal Regulations, Part 50.46, were met.

The following table contains a list of initial conditions and safety actuation setpoints that are modeled in the LOCA analysis.

Parameter Value Core Power 1.02 of 2568 MWt Primary Side Ave. Fluid Temp. 579 0 F RCS Pressure @ HL Tap 2170 psia Total RCS Mass Flow Rate 134x106 bm/hr Low RCS Pressure Rx Trip 1785 psia Reactor Trip Delay 0.6 sec ESAS Low RCS Pressure Trip 1640 psia ESAS Low-Low RCS Pressure Trip 515 psia ESAS (ECCS) Delay Time 67 see - HPI 40 sec - LPI EFW Delay Time 60 sec EFW Actuation On LOOP (Rx Trip)

EFW Flow per SG 200 gpm Manual OTSG Level Control 20 min. after ISCM, fill to ISCM setpoint (26.3 ft)

Automatic OTSG Level Control Fill and maintain with EFW to 50%

OR (20.7 ft)

Main Steam Line Break No specific main steam line break (MSLB) analysis was performed. Relative to increased levels of OTSG tube plugging, the key parameter that can influence the consequences of the MSLB accident is the initial pre-accident SG inventory. The larger initial OTSG inventory will result in the greatest overcooling of the RCS. Typically, the CR-3 plant normally operates with a OTSG level between 65 and 75 percent of the SG operate range, which corresponds to approximately 40,000 to 45,000 lbm per SG. Increasing the level of OTSG tube plugging will increase the level in the boiling region and will increase the mass

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 19 of 29 in this region. The net increase in inventory is less than 5000 lbs to each OTSG and the effective heat transfer area will be proportionally reduced. The MSLB analyses are performed with no OTSG tube plugging (to maximize the heat transfer area) and an OTSG inventory between 55,000 and 56,000 Ibm, which corresponds to approximately 99 percent on the operate range. Since a bounding inventory is modeled in the analysis and the full primary-to-secondary heat transfer area is modeled, the overcooling consequences will be more severe than if OTSG tube plugging were modeled. Therefore, the current analysis, as reported in the Section 14.2.2.1 of the CR-3 FSAR, with will remain bounding. Section 14.2.2.1 defines the initial conditions, safety setpoints, sequences of events and results of the MSLB analysis and this information is not repeated in this response. All of the restrictions and limitations of the approved methods were met for the FSAR analysis.

f. Codes/methodologies used for analyses, approval reference for these codes, and confirmation that all restrictions and limitations included in the approvals of the codes/methodologies were met.

g. Results and plots of Acceptance Criteria parameters.

FPC Response for f. and g.:

DNBR calculations for the four RCP Coastdown, the one RCP Coastdown and the Locked Rotor events were performed using the design assumptions associated with operation with up to 20% OTSG tube plugging. These re-analyses were performed in accordance with methodologies, restrictions, and limitations defined in NRC approved topical reports BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analysis," BAW-10187P-A, "Statistical Core Design for B&W Designed 177 FA Plants,"

and BAW-10156P-A "LYNXT - Core Transient Thermal-Hydraulic Program."

Plots of DNBR versus time for these three events are provided. For all three events, the calculated DNBR remains above the thermal design limit for all time. The current statistically based thermal design limit for CR-3 is 1.40 BWC.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 20 of 29 Figure Q6-1 CR-3 Four Pump Coastdown - 20% OTSG Tube Plugging DNBR versus Time 240 2.30 220 2.10 o 2.00

• 1.90 z

• 1.80 1.70 1.60 1.80 1.40 1.5 2.0 2.5 3.0 3.5 4.0 0.0 0.5 1.0 time, sec.

Figure Q6-2 CR-3 One Pump Coastdown - 20% OTSG Tube Plugging DNBR versus Time m

z 0

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 time, sec.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 21 of 29 Figure Q6-3 CR-3 Locked Rotor - 20% OTSG Tube Plugging DNBR versus Time m

z 0

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 time, sec.

7. NRC Request Withdrawn.

FPC Response Question 7 was resolved prior to issuance of the RAI from the NRC.

8. NRC Requlest:

The Loss of Main Feedwater and Feedwater Line Break events described in the FSAR include an assumption of 20 percent OTSG tube plugging. Did the analysis assume both symmetric and asymmetric tube plugging, and are all Acceptance Criteria for these transients met for these conditions?

FPC Response Only the loss of main feedwater accident was reanalyzed with the higher level of OTSG tube plugging. In the analysis, only symmetric (20 percent) tube plugging was modeled.

The consequences of these accidents are not particularly sensitive to the level or asymmetrical distribution of the OTSG tube plugs. This conclusion is based on a simple comparison of the required emergency feedwater (EFW) flow for the loss of main

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 22 of 29 feedwater accident assuming no OTSG tube plugging and with 20 percent of the tubes plugged. In each case, the required EFW flow rate was 550 gpm. This confirms that, as long as the required EFW flow rate is provided, the effective OTSG heat transfer area is more than adequate to remove core decay heat and the energy addition from the reactor coolant pumps. Therefore, a specific case with asymmetric OTSG tube plugging was not required. In the analysis, all of the event-specific acceptance criteria were met.

9. NRC Request:

Regarding the Small Break Loss of Coolant Accident (LOCA) analyses:

a. The small break LOCA analysis described in the FSAR already includes an assumption of 20 percent OTSG tube plugging. Did the analysis assume both symmetric and asymmetric tube plugging, and are all Acceptance Criteria met for these conditions?

FPC Response:

The analysis modeled asymmetric OTSG tube plugging, specifically 25 percent of the SG tubes in the broken loop and 15 percent of the tubes in the intact loop SG were plugged. Sensitivity studies have shown that asymmetric plugging with the broken loop more heavily plugged is limiting. The analyses were performed in accordance with the NRC-approved BWNT LOCA Evaluation Model (BAW-10192P-A, "BWNT Loss-of Coolant Accident Evaluation Model for Once-Through Steam Generator Plants") and all of the LOCA-related acceptance criteria contained in Title 10 of the Code of Federal Regulations, Part 50.46, were met.

b. The small break LOCA analyses were performed with a maximum of 75 percent of the tubes in the "wetted" region of each OTSG plugged. The licensee suggests that a limit must be established for how many tubes in the "wetted" region can be plugged. Please derme the OTSG "wetted" region, with respect to identification of tubes within this region, and provide information regarding the processes and/or procedures in place to ensure this limit will not be exceeded.

FPC Response:

The SBLOCA analysis progression is sensitive to the number of OTSG tubes that are wetted by the emergency feedwater (EFW) spray. The OTSG model that is used in the LOCA analyses for CR-3 is separated into an EFW wetted region, represented by 10 percent of the total tubes, and an unwetted region that is represented by 90 percent of the tubes. In order to minimize the influence of EFW spray cooling on the transient and to account for increased levels of OTSG tube plugging, the analyses assumed that no more than 75 percent of the tubes in the 10 percent region would be plugged. A review of the CR-3 steam generators indicates that the total number of tubes removed

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 23 of 29 from service through Refueling Outage 12 in SG A was 217 tubes (1.4 percent) and in SG B was 730 tubes (4.7 percent). Since neither of the CR-3 steam generators exceeds 7.5 percent plugging, it is certain that less than 75% of the wetted tubes are currently plugged.

CR-3 procedure SP-305, OTSG Inservice Inspections and Tube Repair Operations, maintains tubesheet maps documenting plugged tubes and requires that reviews be performed to ensure overall and regional plugging limits are not exceeded.

10. NRC Request:

The discussion on the Main Steam Line Break (MSLB) analysis states that "The MSLB design base analyses will remain bounding for the B&W plants with extended tube plugging as long as the error adjusted OTSG level remains below 95% on the operate range, and the plant is not being normally operated with the OTSG aspirator port flooded."

a. This statement appears to be from a generic B&W study, please provide a Reference and a discussion regarding the applicability to Crystal River Unit 3.

FPC Response:

This statement is based on a generic study performed by the B&W Owners Group in Framatome ANP Document 51-5009660-01, "Evaluation of Extended Tube Plugging Limits for the Once-Through Steam Generator." No specific MSLB analysis was performed. Relative to increased levels of OTSG tube plugging, the key parameter that can influence the consequences of the MSLB accident is the initial pre-accident SG inventory. The larger initial OTSG inventory will result in the greatest overcooling of the RCS. Typically, the CR-3 plant normally operates with a OTSG level between 65 and 75 percent of the SG operate range, which corresponds to approximately 40,000 to 45,000 Ibm per SG. Increasing the level of OTSG tube plugging will increase the level in the boiling region and will increase the mass in this region. The net increase in inventory is less than 5000 lbs to each SG. The MSLB analyses are performed with no OTSG tube plugging and an SG inventory between 55,000 and 56,000 Ibm, which corresponds to approximately 99 percent on the operate range. Since a bounding inventory is modeled in the analysis and the full primary-to-secondary heat transfer area is modeled, the overcooling consequences will be more severe. Therefore, the current analysis, as reported in the CR-3 FSAR, with no OTSG tubes plugged will remain bounding. No new analysis will be required to address increased levels of OTSG tube plugging as long as the SG inventory does not exceed what was modeled in the analysis.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 24 of 29

b. Discuss the controls in place at Crystal River Unit 3 to ensure that the MSLB design base analyses remain bounding due to compliance with these constraints.

FPC Response:

Crystal River Technical Specifications, specifically, LCO 3.7.17 and Figure 3.7.17-1 limit the allowable OTSG levels based on the amount of superheat to less than or equal to 96 percent on the operate range.

11. NRC Request:

Regarding the impact of 20 percent OTSG tube plugging on DNB Safety Limits (Figure 6 of the submittal):

a. Please discuss the method of analysis and computer codes used to evaluate the impact of 20 percent OTSG tube plugging on the Statistical Core Design (SCD)

Safety Limits.

FPC Response:

The SCD-Based DNB Safety Limits are generated with the LYNXT computer code using the methods described in Section 6.4 of the NRC approved topical report BAW 10179P-A "Safety Criteria and Methodology for Acceptable Cycle Reload Analysis."

b. Does the "Existing Limits" curve in Figure 6 account for 20 percent OTSG tube plugging?

FPC Response:

The "Existing Limits" curve predates both the SCD analysis method and the 20% tube plugging analyses. However, the purpose of Figure 6 is to demonstrate that the "Existing" curve is conservative, that is 'above and to the left of' the SCD based curves for either 20% tube plugging or no plugging. CR-3 has chosen to retain the conservative "Existing" limit in the plant Technical Specifications even though SCD based analyses, which incorporate the effects of 20% tube plugging, demonstrate that the "Existing" limit is more restrictive.

c. Has CR-3 fully converted to SCD methods such that all transients have been analyzed using SCD? If not, provide justification that the "Existing Limits" curve is bounding.

FPC Response:

CR-3 has fully converted to the SCD analysis method. All DNB limited transients have been reanalyzed with SCD. Figure 6, which represents the steady-state

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 25 of 29 Pressure-Temperature Safety Limits, has also been fully justified using SCD analysis methods which incorporate the effects of 20% tube plugging.

12. NRC Request:

Regarding the impact of 20 percent OTSG tube plugging on the Reactor Protection System (RPS) setpoints:

a. For the High Flux Trip, the MSLB is the most limiting event with respect to overcooling. The submittal states that, "A conservatively low setpoint is modeled such that the MSLB is not limiting for this setpoint." Is this a typographical error, shouldn't this read as, "A conservatively igh setpoint..."? The FSAR MSLB assumes the High Flux Trip is 112 percent, the Technical Specification value is 104.9 percent.

FPC Response:

The statement provided in Licensing Amendment Request (LAR) No. 263, "...a conservative low setpoint...," is misleading as it contradicts what is provided in CR-3 FSAR Table 14-25. FSAR Table 14-25 lists the high flux setpoint used in the analysis as 112 percent of full power. The statement in the LAR was taken from a generic study performed by the B&W Owners Group in Framatome ANP Document 51 5009660-01, "Evaluation of Extended Tube Plugging Limits for the Once-Through Steam Generator." The point that the LAR and the generic B&W Owners Group study is trying to make is that the overcooling events are not limiting relative to setting the high flux trip setpoint. The MSLB accident was identified as the most limiting event in this category because it imposes the most severe overcooling transient on the reactor coolant system. In order to minimize the energy addition from the core, the reactor trip setpoints are modeled in the analysis to ensure that an early reactor trip occurs because this will increase the cooldown rate and can cause the greatest return to power due to subcritical multiplication. Since an early reactor trip is conservative, the overcooling events would require no change to the high flux trip setpoint.

b. Please discuss the methodology used and the results of the calculations performed to justify that the current Variable Low Pressure Trip setpoint remains conservative with up to 20 percent OTSG tube plugging.

FPC Response:

The variable low pressure trip setpoint provides steady-state protection for DNB and essentially represents an adjusted pressure-temperature limit specified by the DNB analysis. The methodology used to develop this trip setpoint is outlined in Section 7.6 of BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," Revision 2. A revised trip setpoint was generated using the NRC-approved

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 26 of 29 methodology, the reduced RCS flow rate associated with 20% OTSG tube plugging and CR-3 specific instrument uncertainties. The revised setpoint was compared to the current setpoint in the CR-3 Improved Technical Specifications (ITS) and it was confirmed that the current trip setpoint was more conservative. Therefore, no change to the setpoint was required. (Note that the current CR-3 variable low pressure trip setpoint was based on earlier, more restrictive, fuel cycle designs and has not been revised to take advantage of the improved analytical methods and tools.)

13. NRC Request:

As a result of higher levels of OTSG tube plugging, RCS pressure will increase slightly, and the licensee has requested a TS change to increase the minimum RCS pressure to 2064 psig. There is no discussion in the submittal related to how this value is calculated; therefore, please provide the basis for this value.

FPC Response:

Based on Framatome's recommendation, CR-3 has requested that the RCS Loop Pressure Limit (SR 3.4.1.1) be increased to 2064 psig. This limit is a measured pressure limit which provides assurance that the nominal 2200 psia core exit pressure is being maintained.

The limit is derived by reducing the nominal core exit pressure by the following three components: Pressure Measurement Uncertainty, psia to psig conversion, and a conservative representation of the AP core exit to pressure tap. As the level of tube plugging in the plant increases, both the flow and the actual value of the "AP core exit to pressure tap" will decrease. Therefore, to ensure that a conservative measured pressure limit is set, the "AP core exit to pressure tap" term must be based on a minimized flow.

The minimized flow associated with 20% tube plugging was used in an evaluation of the existing measured pressure limit. This evaluation indicated that the measurement limit needed to be increased slightly to remain conservative. The nominal core exit pressure has not changed.

14. NRC Request:

The licensee has requested that the following parameters be relocated from the TS to the COLR:

RCS DNB Pressure Limit RCS DNB Temperature Limit RCS DNB Flow Rate Limits RCS Variable Low Pressure Setpoint Equation

a. The proposed change for the RCS pressure limit is not consistent with the temperature and flow changes. The temperature and flow changes maintain

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 27 of 29 absolute limits in the TS and add the cycle-specific limits to the COLR. Why is a limit not being maintained in the TS for RCS pressure also?

FPC Response:

The format for the proposed change was based on amendments approved for other utilities that have moved ITS parameters to the COLR per Generic Letter 88-16 including Arkansas Nuclear One Unit 1 - Amendment 186, dated October 3, 1996 and Byron/Braidwood - Amendment 113/106, dated May 15, 2000. Moving the RCS pressure value to the COLR provides the flexibility to revise the value under the 10 CFR 50.59 process. The change to RCS pressure to account for 20% tube plugging is less than 3 psi. Any future changes to this parameter would also be expected to be very small, but would still require a license amendment if the value remains in the ITS.

The parameters for flow and temperature are more relevant and appropriate to keep in ITS since the plant may actually be operated near those limits if 20% OTSG plugging is required. The value for pressure is an important parameter for analysis but will not alter actual plant operation. Normal RCS operating pressure is approximately 2155 psig and would remain at that value if 20% of the OTSG tubes were plugged. If actual RCS pressure did drop below the limit in the COLR, the required actions of ITS 3.4.1 would be followed.

b. Parameters in the COLR are to be calculated using previously approved NRC methods. Please provide a listing of these methods for the parameters being relocated to the COLR. Current CR-3 TS reference only Topical Report BAW-10179-A. Does this reference cover all methodology for the parameters being relocated to the COLR? Is there an approved topical report that discusses the parameters that can be relocated to an expanded COLR?

FPC Response:

The design basis for "Transient Core Thermal-Hydraulic Analysis" is discussed in Section 6.7 of the NRC approved topical report BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analysis." That section discusses the criteria used to establish the LCOs on RCS Pressure, RCS Hot Leg Temperature and RCS flow rate. With regard to the Variable Low Pressure Setpoint Equation, Section 6.4 of BAW-10179P-A discusses the generation of the RCS DNB Safety Limits (P-T Limits),

while Section 7.6 discusses the methods used to set the Variable Low Pressure Trip.

These sections refer to other approved topical reports, as appropriate, to identify the computer codes and procedures used to analyze these parameters.

c. In relocating the RCS DNB pressure limits to the COLR, the TS limit for three RCP operation is being eliminated. Provide justification for this deletion.

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 28 of 29 FPC Response:

Limit values for both 4 Pump and 3 Pump operation were developed and the limits for both conditions were nearly equal. The small difference between the two values did not warrant having separate limits. The more limiting value (higher RCS pressure) was submitted as the bounding limit for both conditions.

d. Regarding the reason for relocating these parameters to the COLR, the licensee states that this change "will allow the flexibility to utilize the available margins to increase cycle operating margins without the requirement of cycle-specific license amendments." Are the margins that are referred to only those resulting from 20 percent OTSG tube plugging assumptions?

FPC Response:

Yes. Utilizing the parameter limits for 20% OTSG plugging would be limiting for core reload designs when actual plugging is less than 20%. For example, if actual plugging was 5 % of OTSG tubes, a larger value for RCS flow could be used in the analysis compared to the RCS flow value for 20% tube plugging. The ability to utilize the associated margins until actual plugging occurs allows for more economical core designs.

e. Should the RCS Variable Low Pressure function now be listed in the proposed revision to ITS 5.6.2.18a, which lists the parameters documented in the COLR?

FPC Response:

No. ITS Specification 3.3.1 is already listed in ITS 5.6.2.18a. Table 3.3.1-1 is part of Specification 3.3.1. The ALLOWABLE VALUE for Item 8 of this Table, Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE, was previously moved to the COLR. At that time, Specification 3.3.1 was added to the list in ITS 5.6.2.18a. The specific Item of the table was not added to the list. The proposed change for Item 5 of Table 3.3.1-1, the RCS Variable Low Pressure function, is appropriately referenced by ITS 3.3.1. No addition to ITS 5.6.2.18a for the specific table item number is required.

NRC Request For Addition Information dated April 23, 2002 In Attachment F, Section D.4, "Secondary System Performance and Integrity," of your July 24, 2001, letter, Florida Power Corporation (FPC) concluded that the Flow Induced Vibration (FIV) of the tubes will not be significantly affected by a symmetric tube plugging distribution because the tube plugging would result in insignificant increase in dynamic pressure. For an asymmetric plugging distribution

U.S. Nuclear Regulatory Commission Attachment A 3F0602-06 Page 29 of 29 situation, FPC stated that the refinement of FIV analyses could allow for increased feedwater flow under limited power operations. However, the staff noted that it is not clear what effect would increased FIV have, under the asymmetric plugging distribution and full power, on steam generator tube integrity. The staff noted also that potential FIV of once-through steam generator tubes could be subjected to 2-phase flow, vortex shedding, fluid-elastic instability, and turbulence-induced vibration. Therefore, staff requests that FPC provide the results of FIV reassessment calculations or analyses to demonstrate the functional integrity of the steam generator tubes due to increase in steam generator tube plugging up to 20 percent under full power operations for symmetric and worst-case asymmetric plugging distributions. The staff expects that the results will include the fluid-elastic stability margin, input parameters used such as stability constant and viscous damping value, and the impact on wear due to turbulence-induced vibrations.

FPC Response:

The analysis performed for 20% tube plugging did not assume an increase in feedwater flow from that previously analyzed. No other inputs to the flow induced vibration model were changed. Therefore, there was no increase in the calculated flow induced vibration as part of this effort. If 20% of steam generator tubes were actually plugged, CR-3 may not be able to maintain 100% electrical power output with the current feedwater flow. At that time, feedwater flow could be increased to attempt to restore 100% power capability.

The increased feedwater flow would change inputs to the analysis and a revised FIV analysis would be required. The methodology that would be used is the same as that used for evaluating power uprates described in detail in Attachments B and D.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ATTACHMENT B LICENSE AMENDMENT REQUEST #263, REVISION 0 Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging FRA-ANP Engineering Information Record, CR-3 OTSG FIV Margins, document number 51-5000475-01, October 1, 2001, Non-Proprietary

/-U"4U-0 ý1/2UU1)

J FRAMATOME AN P ENGINEERING INFORMATION RECORD Document Identifier 51 - 5000475 - 01 Title CR-3 OTSG FIV MARGINS PREPARED BY: REVIEWED BY:

Name JA BURGESS JR Name HL HASSENPFLUG Signature Date I6JI Ie1 Signature u eate L*/1Zio,,

Technical Manager Statement: Initials Reviewer is Independent.

Remarks:

Rev 01 The purpose of this document is to estimate, from a flow-induced vibration point of view, the margin of safety at the nominal, 100% power (2568 Mwt), corresponding to a maximum reference cross flow velocity at tube location 75-1, equal to 35.9 ft/s [Ref 1]. In addition, the margin of safety for an additional

-1% increase in this cross flow velocity associated with 20% of the tube bundle plugged at 100% power is assessed.

In previous fuel cycles, CR-3 has operated their plant at 2544 Mwt. However, all flow-induced vibration analysis performed for the CR-3 OTSGs was evaluated at 2568 Mwt or the maximum reference cross flow velocity of 35.9 ft/sec at tube location 75-1. The one percent increases in tube bundle cross flow velocities are based upon the revised thermal hydraulic analysis of the CR-3 OTSG performed in Reference

[12]. Due to budget constraints, the revised OTSG secondary side conditions provided in Reference (12] specific to CR-3 could not be evaluated.

Reference [2] is the most recent and comprehensive flow-induced vibration analysis of Although Reference [2] addresses the TMI OTSG, all the results are directly applicable the B&W OTSG tubes.

to the CR-3 OTSG since, the two designs are identical and both TMI and CR-3 have the same power rating and coolant mass flow rate (2568 Mwt).

It is estimated based on the latest CR-3 OTSG tube repair history as of October 1999, the minimum margin of safety occurs in tube 77-26 in the B OTSG. This tube was plugged and stabilized with the flexible stabilizer in May 1992. In the case when this tube is completely severed at the secondary surface of the upper tubesheet, its margin against fluid-elastic instability will be about 8%. Assuming the density remains approximately constant, the reference cross flow velocity can be increased by about 8% without any detrimental effects on the tubes. In the case the secondary density also changes, then the product ('/p).V, where p is the reference secondary mass density and V is the reference cross flow velocity, can be increased by 8% without any detrimental effect on the tubes. The wear rate on the tubes at the top tube support plates will increase by about 43%.

Page 1 of 20

ArRAMATOME AN P 51-5000475-01 RECORD OF REVISIONS Revision Sections Description 00 All Original Release 01 Section 1.0 Complete revision Section 6.0 Revised to incorporate results for 20% tube bundle plugging Section 7 Ref. [6], "32-5000450-01" was "32-5000450-00" Ref. [7], "51-1205898-01" was "51-1205898-00" Ref. [10], "51-1232770-05" was "51-1232770-00" Added Reference [12]

Appendix Added tubes repaired with stabilizers during the October 1999 outage.

Page 2 of 20

fFRAMATOME ANP 51-5000475-01

1. INTRODUCTION The purpose of this document is to estimate, from a flow-induced vibration point of view, the margin of safety at the nominal, 100% power (2568 Mwt) corresponding to a maximum reference cross flow velocity at tube location 75-1 equal to 35.9 ft/s [Ref 1], based on which all flow induced vibration analyses were carried out in the CR-3 type OTSGs. In addition, the margin of safety for an additional -1% increase in this cross flow velocity associated with 20% of the tube bundle plugged at 100% power is assessed. The 1% increases in tube bundle cross flow velocities is based upon the revised thermal hydraulic analysis of the CR-3 OTSG performed in Reference

[12]. Previous FIV analysis of the virgin tube and currently installed stabilizers are evaluated for these cross flow increases using the cross flow velocities in which these stabilizers were most recently evaluated [Reference 2].

2. REVIEW OF FRAMATOME ANP OTSG TUBE FIV ANALYSES Reference [2] is the most recent and comprehensive flow-induced vibration analysis of the B&W OTSG tubes. It covers the virgin tubes, sleeved tubes and tubes with all except the wire rope and the segmented solid rod stabilizers installed into the CR-3 OTSG by Framatome ANP. Although Reference [2] addresses the TMI OTSG, all the results are directly applicable to the CR-3 OTSG since the two designs are identical and both TMI and CR-3 have the same power rating and coolant mass flow rate (5.4E+6 lb/hour). The segmented solid rod and wire rope stabilizers are addressed in References [3] and [4] respectively. Table I summarizes the fluid-elastic stability margins (FSM) and the corresponding modal frequencies, for tube 75-1 at which the reference cross flow gap velocity, V, reaches the maximum value of 35.9 ft/s [1]. Tables 2 and 3 summarize the turbulence-and vortex-induced vibrations for the same tube.

Table 1: Fluid-Elastic Stability Margins at Tube Location 75-1 Vref= 35.9 ft/s (1)

Virgin Tube [2] Severed Tube with Sleeve [2]

Preload 0 lb -250 lb 0 lb -250 lb Mode Hz FSM Hz FSM Hz FSM Hz FSM 1 33.13 1.91 25.23 4.02 33.13 1.91 25.22 4.02 2 36.01 10.7 27.02 5.57 36.01 10.5 27.02 5.56 3 37.89 6.11 30.02 3.16 37.73 2.13 30.00 3.12 4 40.73 3.83 32.20 3.54 38.07 2.08 31.74 2.10 5 42.49 1.83 34.74 1.84 41.42 10.5 33.80 2.43 6 45.19 3.35 37.84 2.24 44.65 13.3 37.16 5.18 7 47.89 6.15 40.98 4.27 47.70 11.3 40.77 7.01 8 52.67 9.95 45.93 6.42 52.58 20.0 45.80 12.4 9 56.17 8.29 49.66 6.12 56.16 19.2 49.66 12.6 10 60.03 15.3 53.99 11.8 60.09 22.5 54.07 17.2 Page 3 of 20

J FRAMATOME ANP 51-5000475-01 Table 1: Continued Severed Tube with Severed Tube with Severed Tube with Severed Tube with Hybrid Stab [2] Flexible Stab [2] Solid Rod Stab [3] Wire Rope Stab [4] (3)

Mode Hz FSM Hz FSM(2) Hz FSM Hz FSM 1 27.43 1.44 22.09 0.76 29.37 1.76 29.4 1.15 2 35.43 4.43 36.20 1.61 33.51 2.51 36.0 1.11 3 36.20 1.61 40.46 12.1 38.28 4.16 40.3 3.33 4 40.53 12.27 40.94 8.37 41.02 7.92 42.2 2.92 5 42.49 8.69 42.50 8.69 45.25 7.93 45.9 4.90 6 47.00 14.52 46.93 15.8 47.77 8.37 48.2 4.01 7 49.88 13.46 50.78 25.4 49.98 10.4 51.4 4.24 8 52.31 14.84 54.08 16.3 53.55 10.5 54.4 5.17 9 56.15 30.52 59.78 22.9 57.29 14.0 59.7 5.68 10 60.89 20.49 65.04 41.4 62.71 14.0 65.2 7.23 Notes:

(1) Minimum values given in bold.

(2) Results for flexible stabilizer at the hypothetical location at 75-1 are obtained from Ref[2],

Appendix A7, before multiplication by 1.2 to take into account bi-stiffness and de-tuning effects.

(3) Results are for cable in the top span with sever at the secondary face of top tubesheet.

Table 2: Turbulence-Induced Vibration at Tube Location 75-1 Max Disp Max Stress Max R Max Disp Max Stress Max R mil rms at psi rms at lb rms at mil rms at psi rms at lb rms at Virgin Tube 0 Preload Virgin Tube -250 lb Preload 5.6 590 0.5 6.3 624 0.44 mid 1 61h top TS top TS mid 16t" top TS top TS Severed Tube with Sleeve 0 Preload Severed Tube with Sleeve -250 lb 5.2 632 0.56 5.3 619 0.48 mid 1 6 "h top TS top TS mid 1 6th top TS top TS Severed Tube with Hybrid, 0 Preload Severed Tube with Flex ,0 Preload (4) 4.9 399 0.47 9.9 1471 0.58 mid 1 6 1h top TS top TS mid mid 16th 1 6 th 1 5 "hTSP Severed Tube with Wire Rope Stab 29.9 2187 2.13 mid 16"' mid 16 "' top TS (4) Results for flexible stabilizer at the hypothetical location at 75-1 are given in Ref[2], Appendix 7.

Page 4 of 20

J FRAMATOME AN P 51-5000475-01 Table 3: Vortex-Induced Vibration at Tube Location 75-1(6)

Max Disp Max Stress Max R Max Disp Max Stress Max R mil 0-p at psi 0-p at lb 0-p at mil 0-p at psi 0-p at lb 0-p at Virgin Tube, 0 Preload Virgin Tube -250 lb Preload 26.2 2680 1.6 30.3 2844 2.0 mid 1 6th top TS top TS mid 16t" top TS top TS Severed Tube with Sleeve 75-1, 0 Preload Severed Tube with Sleeve 75-1, -250 lb 24.0 2903 2.6 24.7 2881 2.1 mid 16' top TS top TS mid 16h top TS top TS Severed Tube with Hybrid 75-1, 0 Preload Severed Tube with Flex, 0 Preload (5) 23.1 1833 2.1 46.1 6817 2.6 mid 16 th top TS top TS mid 16t' mid 16"' 15t' TSP Severed Tube with Wire Rope Stabilizer Not computed Notes:

(5) Results for flexible stabilizer at the hypothetical location at 75-1 are given in Reference [2], Appendix 7.

(6) 0-peak values are equal to sqrt(2)*rms values.

3.0 ASSESSMENT OF MARGINS BASED ON PREVIOUS ANALYSIS RESULTS 3.1 Margin Against Fluid-Elastic Instability (Not Including Flexible Stabilizer)

The first flow-induced vibration to prevent is fluid-elastic instability. As shown in Table 1 and discussed at length in References [2 & 5], the flexible stabilizer should not be used to repair the worst-located tube in the OTSG in cases when this tube is completely severed at the secondary face of the upper tubesheet, as the stability margin for the resulting tube/stabilizer combination would be below 1.0. This is the reason there are exclusion zones for the flexible stabilizer

[5].

Stability for tubes in the CR-3 OTSG with the flexible stabilizers will be discussed in detail in Sections 4 and 5. Excluding the flexible stabilizer, Table 1 shows the minimum FSM is equal to 1.11, for a worst-located tube completely severed at the secondary face of the upper tubesheet and stabilized with the wire rope.

3.2 Margin Against Mid-Span Impact The results of turbulence-induced vibration analyses is summarized in Table 2. The segmented solid rod stabilizer is not included in the analysis. In Reference [3], fluid-elastic stability analysis for the virgin tube and for a severed tube stabilized with the segmented solid rod stabilizer is carried out on the same basis. It is found that a severed tube stabilized with the segmented rod stabilizer actually has higher margin against fluid-elastic instability than the virgin tube.

Page 5 of 20

J FRAMATOME ANP 51-5000475-01 Therefore, it will also have higher margin against turbulence and vortex-induced vibrations. Table 2 shows that the maximum turbulence-induced displacement is 0.0299 in rms which will occur at the middle of the top span in tube 75-1 stabilized with a wire rope stabilizer [4]. This is about E the half-gap clearance between tubes. From Reference [11], the probability to exceed the 4 sigma4 point is 6.17E-5. Following the same discussion as in Reference [2], the probability that this tube would hit its neighbor is 6.17E-5^2. At 30 Hz and over 40 years, two neighboring tubes both stabilized with wire ropes will hit each other a total of about 144 times. This is an acceptable risk.

However, this also shows that there is no margin of safety in this case.

Vortex-induced vibration is not addressed in Reference [4]. However, by extrapolating the results from Reference [2] for the flexible stabilizer, one can readily deduce that the maximum vortex induced displacement at the middle of the top span for the wire rope stabilizer is probably around 0.0299 x (0.0326/0.0099) = 0.098 in. 0-peak, compared with the half tube-tube clearance of 0.125 in. Thus, there is 28% margin (13% in main feedwater mass flow rate) of safety against tube-to tube impacting caused by vortex-induced vibration even in the worst case of a wire-rope stabilized tube.

3.3 Margin Against Fatigue Table 2 shows that the maximum bending stress of 2187 psi rms would occur in the middle of the top span of tube 75-1, if the tube is completely severed at the secondary surface of the top tubesheet and stabilized with the wire rope. As is discussed in Section 7.2 of Reference [2], the rms endurance limit, as derived from the 0-peak ASME fatigue curve "C", is 3450 psi rms. Thus even under the most conservative assumptions, there is more than 57% margin against fatigue due to turbulence-induced vibration.

Using the same reasoning as in Section 3.1, one can estimate that the corresponding vortex induced stress in this tube would be about 4821 x (2187/1471) = 7169 psi 0-peak. This is far below the ASME endurance limit (more than 10,000 psi 0-peak) for austenitic steel. The margin of safety for vortex-induced fatigue is at least 40%.

3.4 Margin Against Wear Wear of the tubes against the support plates is not addressed in References [2] through [4], but is addressed in Reference [6]. Based on non-linear, time domain analyses, it is estimated that under normal operational conditions, the worst-located virgin tube will lose approximately 0.0054 inch of its wall thickness after 5 years of effective full power operation. The wear rates for tubes with sleeves and stabilizers are not addressed in References [2-4]. However, rough estimates on the increase in the wear rate as a function of main feed water flow rate can be made using the following reasonable assumptions: (1) For small differences in the mean cross flow gap velocities, the wear rate varies approximately as the product of the mid-span tube displacement and the reaction force at the TSPs. (2) Again for small differences in the mean cross flow gap velocities, both the reaction force and the mid-span tube displacement vary as the forcing function which varies as the square of the cross flow velocity. Thus, the tube wear rate varies approximately as the 4 th power of the main feedwater flow rate. For a virgin tube, the rate of wall thinning is only 0.0054 inch for 5 EFPY of operation [6]. Therefore, any small increases in the main feedwater flow rate will not have a detrimental effect on tube wear.

Page 6 of 20

ArRAMATOM E AN P 51-5000475-01 The wear on the virgin tube should bound the wear on the sleeved tube as the latter has the entire tube wall thickness to wear out before feeling any harmful effect. The stabilized tubes have been plugged and removed from service. Therefore, wear is not a concern for the stabilized tubes.

4. REVIEW OF CR-3 OTSG TUBE REPAIR HISTORY Table 1 shows that a tube that is completely severed at the secondary surface of the upper tubesheet and stabilized with the flexible stabilizer will be unstable, with a stability margin of only 0.76. This increases to 0.91 if, following the reasoning given in Reference [2], bi-stiffness and de tuning effect each gives the flexible stabilizer an additional 10% of margin. This is still under the stability threshold of 1.0. This is why there are exclusion zones in the OTSG within which use of the flexible stabilizer is restricted [5]. In addition, it is shown in Section 3.2 that a cluster of worst-located, completely severed tubes repaired with wire rope stabilizers will have no margin against tube-to-tube impacts due to turbulence. Thus, the minimum margin against flow-induced vibration is bounded by tubes with the flexible or wire rope stabilizers in the CR-3 OTSGs.

A comprehensive tube repair history of CR-3 is given in the Appendix. Listed are the location, date of installation and types of stabilizer or sleeve of each of the tube ever repaired by Framatome ANP. Cross referencing the Appendix and the reference cross flow velocity contour plots in Reference [1], which are reproduced in figure 1, shows that one or more of the following tubes are limiting: 72-5; 73-1, 73-19, 74-22,77-12, 77-26, 78-13. The first tube (in the B OTSG) has a wire rope stabilizer while the rest have flexible stabilizers,

5. ASSESSMENT OF MARGINS FOR TUBES WITH WIRE ROPE AND FLEXIBLE STABILIZERS 5.1 Wire Rope Stabilized Tube The most marginal tube with the wire rope stabilizer is tube 72-5 in the B OTSG at which the reference cross flow velocity is, from Figure 1, 27 ft/sec, compared with 35.9 ft/sec at location 75

1. Thus, the margin at this worst location is 47% for the 100% power operating condition and 46% for the 100% power and 20% plugging.

5.2 Tubes with Flexible Stabilizers The reference cross flow velocity V for each of the tubes with the flexible stabilizer in the CR-3 OTSGs can be obtained from Figure 1. This is reproduced in the third column of Table 4. From these, the fluid-elastic stability margins (FSM) for each of them can be computed:

FSM = 0.91 x (35.9/V)

Page 7 of 20

A' .RAMATOME AMP 51-5000475-01 Figure 1: Reference Cross Flow Gap Velocities in the Lane Region of the OTSG

.RED ZONE = 34 tubes, use of laminated stabilizers not recommended PINK ZONE= 50 tubes, use of two laminated stabilizers side by side not recommended.

Page 8 of 20

JFRAMATOME ANP 51-5000475-01 Table 4: FSM for the Most Marginal Tubes with Flex Stabilizer Tube Location Date Installed Reference V FSM ft/sec 73-1 May 92 31.5 1.04 73-19 May 92 21 1.56 74-22 May 92 <21 >1.56 77-12 May 92 31 1.05 77-26 May 92 30 1.09 78-13 May 92 <21 >1.56 All others 1990-1992 <26.7 >1.22

• Installed for precautionary measure. Tube flaw indication was at noise floor level of the eddy current sensor.

Review of Table 4 shows that there is a minimum of 4% margin against fluid-elastic instability for tubes with the flexible stabilizer, and that this minimum margin occurs at tube 73-1. Reference [7]

shows that the reason for plugging this tube was a small signal to noise (S/N) outside diameter (OD) indication between the 13th and 14th TSP. Also a UTS "C-type" indication was reported at the upper tubesheet secondary face (UTSF) [8]. The C-type indication is an anomalous, non defect signal. The tube was stabilized in accordance with Reference [10] because it is located within the defined lane/wedge region, where secondary side cross flow is highest. Since this tube did not have any flaw indications at either the UTS or the 15th TSP when it was plugged, it is not considered necessary to assume that the tube is severed when evaluating the acceptability of the stabilizer at this location. If tube 73-1 is not completely severed, then its stiffness will be higher than that of the virgin tube. This, together with the added damping due to the flexible stabilizer, will result in a margin against fluid-elastic instability higher than that of the virgin tube.

The next tube with the lowest margin against fluid-elastic instability is tube 77-12 in the B-OTSG.

This tube was plugged in the 5/92 outage with a rolled plug and flexible stabilizer [7] due to two ECT indications located 15.5" and 18.0" above the UTS secondary side face. No additional indications were reported in the tube [8]. The tube was stabilized in accordance with Reference

[10] due to its location along the lane in a high cross flow area. The two UTS indications are located deep enough into the tubesheet that the tube would be considered to be fixed even if these flaws were to eventually propagate. Since there were no indications of degradation at the UTS or the 15th TSP, it is not necessary to assume that the tube is severed when evaluating the acceptability of the stabilizer at this location. Following the same reasoning as for tube 73-1, this tube also has a margin against fluid-elastic instability higher than that of a virgin tube.

6. CONCLUSION Based on the above analysis, it is concluded that, the CR-3 OTSG tubes presently (or latest outage, October 1999) have a minimum of 9% margin against detrimental results due to flow-induced vibration for the 100% power, 0% plugging condition and 8% margin for the 100% power with 20% plugging condition. The limiting tube is No.77-26 in the B OTSG, which has been plugged and stabilized with a flexible stabilizer in May 1992. The limiting FIV mechanism is fluid-elastic Page 9 of 20

Y rFRAMATOME AN P 51-5000475-01 instability. Since the margin against fluid-elastic instability is inversely proportional to the square root of the mean secondary fluid density multiplied by the mean cross flow velocity V, the product (4p)V can be increased by 8% without long term detrimental effects on any of the tubes. It is estimated that at this higher mass flow rate, the increase in the wear rates on all the tubes due to tube-to-TSP interaction is about 43%.

7. REFERENCE

1. L.E.Johnson, "Velocity and Density in 177FA OTSG", 51-1171541-00, 88/7

2. M.K. Au-Yang, "Flow-Induced Vibration Analysis of TMI OTSG Tubes due to Power Uprate,"

FTI Doc 32-1257514-01, June 1997.

3. B.Brenneman, "Fluid Elastic Instability Margins for 177 OTSG Tubes with Various Stabilizers",

32-1143925-00, July 1983.

4. B. Brenneman, "OTSG Cable Stabilizer-FIV," Framatome ANP Doc 32-1225188-00, Sept 93.

5. M.K. Au-Yang, "Exclusion Zone for 2-Span Laminated Stabilizer," Framatome ANP Doc 51 1168535-00, April, 1989.

6. M.K. Au-Yang, "ANO-1 OTSG Auxiliary Feed Water Capacity", Framatome ANP Doc 32 5000450-01, October 1997.

7. Framatome ANP Doc. 51-1205898-01, "CR-3 5/92 Plug Removal, Roll Plug", dated July 1994.

8. "Eddy Current Examination Report for Crystal River Unit No. 3, May 1992 Refueling Outage No.

8", NRC Docket No. 50-302, July 1992.

9. Framatome ANP Document 51-5000226-00, "Stabilization Criteria for CR-3 6/97 Inspection",

dated June 1997.

10. Framatome ANP Document 51-1232770-05, "OTSG Tube Stabilization Criteria",

dated January 2001.

11. E.S. Pearson, H.O. Hartley, "Biometrika Tables", Cambridge University Press, 1966.

12. Framatome ANP Document 32-5012972-00, "CR-3 Power Uprate Operating Conditions",

dated 6/2001.

Page 10 of 20

J RAMATOME ANP 51-5000475-01 APPENDIX List of stabilizers installed into the CR-3 OTSGs (thur 10/99 outage)

ROW COL GEN LEG CODE TYPE HEAT# SIN MATL MFG DATE INSTALLED DATE REMO 4 39 S/G A UTE STB WIRE ROPE CF36-2 51 1690 FTI Jul 24 1997 4 39 S/G A UTE STB WIRE ROPE CF36-2 59 1690 FTI Jul 24 1997 4 40 SIGA UTE STB FLEX NX5398 37 1600 BW Mar 1 1990 7 28 S/G A IJTE STB FLEX 398-730 117 1600 BW May 1 1992 25 91 S/G A UTE STB FLEX NX5398 20 1600 8W Mar 1 1990 26 91 SIG A UTE STB FLEX 398-730 135 1600 BW May 1 1992 27 91 S/G A LTE STB WIRE ROPE 9654-1 173 1690 BW Apr 28 1994 27 91 S/G A UTE STS WIRE ROPE 9654-1 182 1690 8W Apr 27 1994 32 107 S/G A UTE STB FLEX 398-730 90 1600 aW May 1 1992 41 15 SIGA UTE STB WIRE ROPE 9654-1 153 BW Mar18 1996 63 129 S/G A LTE STB FLEX NX3624- 233 1600 BW Mar 1 1990 70 1 S/G A SLV ROLLED 764371 732 1690 BW Apr 26 1994 71 1 S/G A SLV ROLLED 764371 699 1690 6W Apr 26 1994 71 2 S/GA SLV ROLLED 764371 739 1690 BW Apr 26 1994 71 3 S/G A SLV ROLLED 764371 758 1690 8W Apr 26 1994 71 59 S/G A UTE STB WIRE ROPE CF36-2 76 1690 FTI Jul 24 1997 71 61 S/GA UTE STB FLEX 398-730 128 1600 BW May 1 1992 72 1 S/G A SLV ROLLED 764371 191 1690 BW Apr 26 1994 72 2 S/G A SLV ROLLED 764371 783 1690 BW Apr 26 1994 72 3 S/G A SLV ROLLED 764371 719 1690 8W Apr 26 1994 72 4 S/G A SLV ROLLED 764371 738 1690 8W Apr 26 1994 72 16 S/G A UTE STB WIRE ROPE CF36-2 55 1690 FTI Jul 24 1997 72 28 S/G A UTE STB WIRE ROPE CF36-2 51 1690 FTI Jul 24 1997 72 30 S/G A UTE STB WIRE ROPE CF36-2 62 1690 FTI Jul 24 1997 72 67 SIG A UTE STB FLEX 398-730 105 1600 BW May 11992 73 1 S/G A SLV ROLLED 764371 782 1690 8W Apr 26 1994 73 2 S/G A "SLV ROLLED 764371 780 1690 BW Apr 26 1994 73 3 S/GA SLV ROLLED 764371 788 1690 BW Apr 26 1994 73 4 S/G A SLV ROLLED 764371 718 1690 BW. Apr 26 1994 73 5 S/GA SLV ROLLED 764371 717 1690 8W Apr 26 1994 73 6 S/G A SLV ROLLED 764371 737 1690 BW Apr 26 1994 73 12 S/GA UTE STB WIRE ROPE 9654-1 159 BW Mar 18 1996 73 63 SIG A UTE ST8 FLEX 398-730 74 1600 BW May 1 1992 73 66 SIG A UTE STB WIRE ROPE CF36-2 39 1690 FTI Jul 25 1997 74 1 S/G A SLV ROLLED 764371 233 1690 8W Apr 26 1994 74 2 SIG A SLV ROLLED 764371 789 1690 BW Apr 26 1994 74 3 S/G A SLV ROLLED 764371 797 1690 BW Apr 26 1994 74 4 S/G A SLV ROLLED 764371 773 1690 8W Apr26 1994 74 5 S/G A SLV ROLLED 764371 757 1690 BW Apr 26 1994 74 6 S/G A SLV ROLLED 764371 795 1690 8W Apr 26 1994 74 7 S/G A SLV ROLLED 764371 754 1690 BW Apr 26 1994 74 8 S/G A SLV ROLLED 764371 759 1690 BW Apr 26 1994 74 9 S/G A SLV ROLLED 764371 779 1690 BW Apr 26 1994 74 10 S/G A SLV ROLLED 764371 720 1690 8W Apr 26 1994 74 11 S/G A SLV ROLLED 764371 747 1690 BW Apr 26 1994 74 12 S/G A SLV ROLLED 764371 752 1690 8W Apr 26 1994 74 13 S/G A SLV ROLLED 764371 716 1690 BW Apr 26 1994 74 14 S/G A SLV ROLLED 764371 707 1690 BW Apr 26 1994 74 15 S/G A SLV ROLLED 764371 715 1690 8W Apr 26 1994 74 16 S/G A SLV ROLLED 764371 756 1690 BW Apr 26 1994 74 17 S/G A SLV ROLLED 764371 793 1690 BW Apr 26 1994 Page 11 of 20

fFRAMATOME ANP 51-5000475-01 74 18 S/G A SLV ROLLED 764371 775 1690 BW Apr26 1994 74 19 S/G A SLV ROLLED 764371 799 1690 8W Apr 26 1994 74 20 S/G A SLV ROLLED 764371 412 1690 BW Apr 26 1994 74 21 S/G A SLV ROLLED 764371 708 1690 BW Apr 26 1994 74 22 S/G A SLV ROLLED 764371 776 1690 BW Apr 26 1994 74 23 S/G A SLV ROLLED 764371 505 1690 8W Apr 26 1994 74 24 S/G A SLV ROLLED 764371 787 1690 BW Apr 26 1994 74 25 S/G A SLV ROLLED 764371 476 1690 sW Apr 26 1994 74 26 S/G A SLV ROLLED 764371 781 1690 BW Apr 26 1994 74 27 S/G A SLV ROLLED 764371 785 1690 BW Apr 26 1994 74 28 S/G A SLV ROLLED 764371 751 1690 BW Apr 26 1994 74 29 S/G A SLV ROLLED 764371 741 1690 BW Apr 26 1994 74 30 S/G A UTE STB FLEX 398 114 1600 BW May 1 1992 74 58 SIG A UTE STB FLEX 398-730 42 1600 BW May 1 1992 74 63 S/GA UTE STB WIRE ROPE CF36-2 81 1690 FTI Jul 24 1997 75 1 S/G A SLV ROLLED 764371 428 1690 BW Apr 26 1994 75 2 S/GA SLV ROLLED 764371 711 1690 BW Apr 26 1994 75 3 S/G A SLV ROLLED 764371 792 1690 BW Apr 26 1994 75 4 S/G A SLV ROLLED 764371 713 1690 8W Apr 26 1994 75 5 S/G A SLV ROLLED 764371 786 1690 8W Apr 26 1994 75 6 S/G A SLV ROLLED 764371 143 1690 8W Apr 26 1994 75 7 S/G A SLV ROLLED 764371 493 1690 BW Apr 26 1994 75 8 S/G A SLV ROLLED 764371 733 1690 BW Apr 26 1994 75 9 S/G A SLV ROLLED 764371 372 1690 BW Apr 26 1994 75 10 S/G A SLV ROLLED 764371 167 1690 BW Apr 26 1994 75 11 S/GA SLV ROLLED 764371 478 1690 BW Apr 26 1994 75 12 S/G A SLV ROLLED 764371 677 1690 BW Apr 26 1994 75 13 S/G A SLV ROLLED 764371 165 1690 8W Apr 26 1994 75 14 S/G A SLV ROLLED 764371 467 1690 8W Apr 26 1994 75 15 SIG A SLV ROLLED 764371 210 1690 BW Apr 26 1994 75 16 S/G A SLV ROLLED 764371 361 1690 BW Apr 26 1994 75 17 S/G A SLV ROLLED 764371 366 1690 BW Apr 26 1994 75 18 S/G A SLV ROLLED 764371 190 1690 8W Apr26 1994 75 19 S/G A SLV ROLLED 764371 468 1690 BW Apr 26 1994 75 20 S/G A SLV ROLLED 764371 339 1690 BW Apr 26 1994 75 21 S/G A S LV ROLLED 764371 174 1690 BW Apr 26 1994 75 22 S/G A SLV ROLLED 764371 145 1690 BW Apr 26 1994 75 23 S/G A SLV ROLLED 764371 144 1690 BW Apr 26 1994 75 24 S/G A SLV ROLLED 764371 188 1690 BW Apr 26 1994 75 25 S/G A SLV ROLLED 764371 186 1690 BW Apr26 1994 75 26 S/G A SLV ROLLED 764371 455 1690 BW Apr26 1994 75 27 S/G A SLV ROLLED 764371 153 1690 BW Apr26 1994 75 28 S/G A SLV ROLLED 764371 481 1690 BW Apr 26 1994 75 29 S/G A SLV ROLLED 764371 192 1690 BW Apr 26 1994 75 30 S/G A SLV ROLLED 764371 157 1690 BW Apr 26 1994 75 31 S/G A SLV ROLLED 764371 162 1690 BW Apr 26 1994 75 32 S/G A "SLV ROLLED 764371 346 1690 BW Apr 26 1994 75 33 S/G A SLV ROLLED 764371 152 1690 BW Apr 26 1994 75 34 S/G A SLV ROLLED 764371 363 1690 BW Apr 26 1994 75 35 S/G A SLV ROLLED 764371 508 1690 BW Apr 26 1994 75 36 S/G A SLV ROLLED 764371 499 1690 BW Apr 26 1994 75 37 S/G A SLV ROLLED 764371 370 1690 BW Apr 26 1994 75 38 S/G A SLV ROLLED 764371 155 1690 8W Apr 26 1994 75 39 S/G A SLV ROLLED 764371 486 1690 BW Apr 26 1994 75 40 S/G A SLV ROLLED 764371 206 1690 BW Apr 26 1994 75 41 S/G A SLV ROLLED 764371 198 1690 BW Apr 26 1994 75 52 S/G A UTE ST8 FLEX 398-730 31 1600 BW May 1 1992 75 60 S/G A UTE STB WIRE ROPE CF36-2 66 1690 FTI Jul 25 1997 Page 12 of 20

fFRAMATOME AN P 51-5000475-01 77 1 S/G A SLV ROLLED 764371 357 1690 BW Apr 26 1994 77 2 S/GA SLV ROLLED 764371 460 1690 8W Apr 26 1994 77 3 S/G A UTE STB FLEX NX5398 28 1600 BW Mar 1 1990 77 3 S/G A UTE STB SGMT NX3232G 404 1600 BW Mar 1 1990 77 3 SIG A UTE STB SGMT NX3232G 471 1600 BW Mar 1 1990 77 3 S/G A UTE STB SGMT NX3232G 489 1600 BW Mar 1 1990 77 3 S/G A UTE STB SGMT NX3232G 499 1600 BW Mar 1 1990 77 4 S/G A UTE STB FLEX NX5398 13 1600 BW Mar 1 1990 77 4 S/G A UTE STB SGMT NX3232G 418 1600 BW Mar 1 1990 77 4 S/G A UTE STB SGMT NX3232G 446 1600 BW Mar 1 1990 77 4 S/G A UTE *.STB SGMT NX3232G 461 1600 BW Mar 1 1990 77 4 S/G A UTE STB SGMT NX3232G 492 1600 sW Mar 1 1990 77 5 SIG A UTE STB HYBRID 398-730 78 1600 BW May 1 1992 77 5 S/GA UTE STB HYBRID NX3232G 523 1600 BW May 1 1992 77 5 S/GA UTE STS HYBRID NX3232G 549 1600 BW May 1 1992 77 5 S/GA UTE STB HYBRID NX3232G 554 1600 BW May 1 1992 77 5 S/G A UTE ST8 HYBRID NX3232G 561 1600 BW May 1 1992 77 6 S/G A SLV ROLLED 764371 482 1690 BW Apr 26 1994 77 7 S/G A SLV ROLLED 764371 500 1690 BW Apr 26 1994 77 8 SIG A SLV ROLLED 764371 342 1690 BW Apr 26 1994 77 9 S/G A SLV ROLLED 764371 309 1690 8W Apr 26 1994 77 10 S/G A SLV ROLLED 764371 497 1690 BW Apr 26 1994 77 11 S/GA SLV ROLLED 764371 683 1690 BW Apr 26 1994 77 12 SIG A SLV ROLLED 764371 682 1690 BW Apr 26 1994 77 13 S/GA SLV ROLLED 764371 473 1690 BW Apr 26 1994 77 14 S/G A SLV ROLLED 764371 470 1690 8W Apr 26 1994 77 15 S/GA SLV ROLLED 764371 343 1690 8W Apr 26 1994 77 16 S/G A SLV ROLLED 764371 284 1690 8W Apr 26 1994 77 17 S/G A SLV ROLLED 764371 489 1690 BW Apr 26 1994 77 18 S/G A SLV ROLLED 764371 348 1690 BW Apr 26 1994 77 19 S/G A SLV ROLLED 764371 161 1690 BW Apr 26 1994 77 20 S/G A SLV ROLLED 764371 365 1690 8W Apr 26 1994 77 21 S/G A SLV ROLLED 764371 377 1690 8W Apr 26 1994 77 22 S/G A SLV ROLLED 764371 496 1690 8W Apr 26 1994 77 23 S/G A .SLV ROLLED 764371 350 1690 8W Apr 26 1994 77 24 S/G A SLV ROLLED 764371 340 1690 8W Apr 26 1994 77 25 S/G A SLV ROLLED 764371 387 1690 sW Apr 26 1994 77 26 S/G A UTE ST8 FLEX 398-730 44 1600 BW May 1 1992 77 27 S/G A SLV ROLLED 764371 353 1690 BW Apr 26 1994 77 28 S/G A SLV ROLLED 764371 338 1690 BW Apr 26 1994 77 29 SIG A SLV ROLLED 764371 488 1690 8W Apr 26 1994 77 30 S/G A SLV ROLLED 764371 762 1690 BW Apr 26 1994 77 31 S/G A SLV ROLLED 764371 735 1690 BW Apr 26 1994 77 32 S/G A SLV ROLLED 764371 761 1690 BW Apr 26 1994 77 33 S/G A SLV ROLLED 764371 778 1690 BW Apr 26 1994 77 34 S/G A SLV ROLLED 764371 791 1690 BW Apr 26 1994 77 35 S/G A SLV ROLLED 764371 784 1690 BW Apr 26 1994 77 36 S/G A SLV ROLLED 764371 796 1690 BW Apr 26 1994 77 38 S/G A SLV ROLLED 764371 749 1690 BW Apr 26 1994 77 39 S/G A SLV ROLLED 764371 195 1690 8W Apr 26 1994 77 40 S/G A SLV ROLLED 764371 736 1690 BW Apr 26 1994 77 41 S/G A SLV ROLLED 764371 504 1690 BW Apr 26 1994 77 43 S/G A UTE STB FLEX 398-730 55 1600 8W May 1 1992 77 53 S/G A UTE STB WIRE ROPE CF36-2 98 1690 FTI Jul 24 1997 77 61 S/G A UTE STB WIRE ROPE CF36-2 94 1690 FTI Jul 25 1997 77 62 S/G A UTE STS WIRE ROPE CF36-2 93 1690 FTI Jul 25 1997 78 1 S/G A SLV ROLLED 764371 396 1690 BW Apr 26 1994 78 2 S/G A SLV ROLLED 764371 347 1690 BW Apr 26 1994 Page 13 of 20

fFRAMATOME ANP 51-5000475-01 78 3 S/G A SLV ROLLED 764371 385 1690 BW Apr 26 1994 78 4 S/G A SLV ROLLED 764371 490 1690 8W Apr 26 1994 78 5 S/G A SLV ROLLED 764371 368 1690 8W Apr26 1994 78 6 S/GA SLV ROLLED 764371 472 1690 BW Apr 26 1994 78 7 S/G A SLV ROLLED 764371 395 1690 BW Apr 26 1994 78 8 S/G A SLV ROLLED 764371 397 1690 BW Apr 26 1994 78 9 S/G A SLV ROLLED 764371 388 1690 BW Apr 26 1994 78 10 S/G A SLV ROLLED 764371 362 1690 sW Apr 26 1994 78 11 S/G A SLV ROLLED 764371 313 1690 BW Apr 26 1994 78 12 S/G A SLV ROLLED 764371 394 1690 BW Apr 26 1994 78 13 S/GA SLV ROLLED 764371 352 1690 BW Apr 26 1994 78 14 S/G A SLV ROLLED 764371 303 1690 sW Apr 26 1994 78 15 SIGA SLV ROLLED 764371 228 1690 sW Apr 26 1994 76 16 S/G A SLV ROLLED 764371 384 1690 BW Apr 26 1994 78 17 S/G A SLV ROLLED 764371 389 1690 BW Apr 26 1994 78 18 S/G A SLV ROLLED 764371 330 1690 BW Apr26 1994 78 19 S/GA SLV ROLLED 764371 349 1690 BW Apr 26 1994 78 20 SIG A SLV ROLLED 764371 506 1690 BW Apr 26 1994 78 21 S/G A SLV ROLLED 764371 479 1690 BW Apr 26 1994 78 22 S/G A SLV ROLLED 764371 341 1690 BW Apr 26 1994 78 23 S/G A SLV ROLLED 764371 458 1690 BW Apr 26 1994 78 24 S/G A SLV ROLLED 764371 296 1690 BW Apr26 1994 78 25 S/G A SLV ROLLED 764371 483 1690 BW Apr 26 1994 78 26 S/G A SLV ROLLED 764371 380 1690 BW Apr26 1994 78 27 S/G A SLV ROLLED 764371 317 1690 BW Apr 26 1994 78 28 S/G A SLV ROLLED 764371 355 1690 BW Apr 26 1994 78 29 S/G A SLV ROLLED 764371 376 1690 BW Apr 26 1994 78 54 S/G A UTE STB WIRE ROPE CF36-2 82 1690 FTI Jul 24 1997 78 103 S/G A UTE STB WIRE ROPE CF36-2 60 1690 "FTI Jul 25 1997 79 1 S/G A SLV ROLLED 764371 373 1690 sW Apr 26 1994 79 2 S/G A SLV ROLLED 764371 462 1690 BW Apr 26 1994 79 3 SiG A SLV ROLLED 764371 495 1690 BW Apr 26 1994 79 4 S/G A SLV ROLLED 764371 405 1690 BW Apr 26 1994 79 5 S/G A SLV ROLLED 764371 692 1690 SW Apr 26 1994 79 6 S/G A SLV ROLLED 764371 425 1690 BW Apr 26 1994 79 28 S/G A UTE STB FLEX 398 97 1600 BW May 1 1992 79 30 S/G A UTE STB FLEX 398-730 84 1600 BW May 1 1992 79 34 S/G A UTE ST8 FLEX 398 143 1600 8W May 1 1992 79 36 SIG A UTE STB WIRE ROPE CF36-2 28 FTI 1690 Jul24 1997 80 1 S/G A SLV ROLLED 764371 689 BW 1690 Apr 26 1994 80 2 S/G A SLV ROLLED 764371 464 1690 BW Apr 26 1994 80 3 SIG A SLV ROLLED 764371 465 1690 8W Apr 26 1994 80 4 S/G A SLV ROLLED 764371 423 1690 BW Apr 26 1994 80 20 S/G A UTE STB WIRE ROPE CF36-2 40 1690 FTI Jul 24 1997 80 26 SIG A UTE STB WIRE ROPE CF36-2 75 1690 FTI Jul 24 1997 80 28 S/G A UTE STB WIRE ROPE CF36-2 68 1690 FTI Jul 24 1997 80 33 S/G A UTE 5TB WIRE ROPE CF36-2 50 1690 FTI Jul 24 1997 81 1 S/G A SLV ROLLED 764371 378 1690 BW Apr 26 1994 81 2 S/G A SLV ROLLED 764371 351 1690 BW Apr26 1994 81 3 S/G A SLV ROLLED 764371 466 1690 BW Apr 26 1994 81 36 S/G A UTE STB WIRE ROPE CF36-2 78 1690 FTI Jul 24 1997 82 1 S/G A SLV ROLLED 764371 414 1690 BW Apr 26 1994 86 96 S/G A UTE STB WIRE ROPE 9654-1 169 BW Mar 18 1996 88 1 S/G A UTE ST8 SGMT  ?  ? 1600 8W Nov 1 1981 111 68 S/GA UTE STB WIRE ROPE 9654-1 171 BW Mar 18 1996 126 71 S/G A UTE STB WIRE ROPE CF36-2 27 1690 FTI Jul 24 1997 141 57 S/GA UTE STB WIRE ROPE 9654-1 183 BW Mar 18 1996 43 63 S/G A LTE STB WIRE ROPE 9654-1 484 1690 FTI Oct 31 1999 Page 14 of 20

A RAMATOME AN P 51-5000475-01 44 64 SIG A LTE STB WIRE ROPE 8245 274 1690 FTI Oct 31 1999 45 54 SIG A LTE STB WIRE ROPE 8245 304 1690 FTI Oct 31 1999 45 55 SIG A LTE STB WIRE ROPE 8245 301 1690 FTI Oct31 1999 46 62 SIG A LTE STB WIRE ROPE 8245 278 1690 FTI Oct 31 1999 48 54 S/G A LTE STB WIRE ROPE 9654-1 412 1690 FTI Oct 31 1999 49 62 S/G A LTE STB WIRE ROPE 8245 271 1690 FTI Oct 31 1999 54 89 S/G A 55 LTE STB WIRE ROPE "8245 281 1690 FTI Oct 31 1999 89 S/G A LTE STB WIRE ROPE 8245 285 1690 FTI Oct 31 1999 56 89 SIG A LTE STB WIRE ROPE 8245 302 1690 FTI Oct 31 1999 58 90 S/G A LTE STB WIRE ROPE 8245 279 1690 FTI Oct 31 1999 58 92 S/G A LTE STB WIRE ROPE 8245 299 1690 FTI Oct31 1999 79 7 S/G A UTE STB WIRE ROPE 8245 292 1690 FTI Oct 31 1999 103 49 SIG A LTE STB WIRE ROPE 8245 318 1690 FTI Oct 31 1999 104 46 S/G A LTE STB WIRE ROPE 8245 306 1690 FTI Oct 31 1999 104 52 S/G A LTE STB WIRE ROPE CF36-2 92 1690 FTI Oct 31 1999 105 47 S/G A LTE STB WIRE ROPE 8245 295 1690 FTI Oct 31 1999 105 48 S/G A LTE STB WIRE ROPE 8245 309 1690 FTI Oct31 1999 105 51 S/GA LTE STS STB WIRE ROPE CF36-2 42 1690 FTI Oct 31 1999 106 51 S/GA LTE WIRE ROPE CF36-2 58 1690 FTI Oct 31 1999 106 52 S/G A LTE STB WIRE ROPE 8245 469 1690 FTI Oct 31 1999 107 53 S/G A LTE STB WIRE ROPE 8245 426 1690 FTI Oct 31 1999 107 55 S/G A LTE STB WIRE ROPE CF36-2 1 1690 FTI Oct 31 1999 107 60 S/G A LTE STB WIRE ROPE 8245 599 1690 FTI Oct 31 1999 108 49 S/G A LTE STB WIRE ROPE CF36-2 37 1690 FTI Oct 31 1999 108 54 S/G A LTE STB WIRE ROPE 8245 532 1690 FTI Oct31 1999 108 55 S/G A LTE STB WIRE ROPE 8245 456 1690 FTI Oct31 1999 109 48 S/G A LTE STB WIRE ROPE CF36-2 46 1690 FTI Oct 31 1999 109 51 S/G A LTE STIB ST8 WIRE ROPE CF36-2 70 1690 FTI Oct 31 1999 109 52 S/G A LTE WIRE ROPE STB 8245 440 1690 FTI Oct 31 1999 109 54 S/G A LTE WIRE ROPE STB 8245 462 1690 FTI Oct 31 1999 109 58 S/GA LTE WIRE ROPE STB CF36-2 47 1690 FTI Oct 31 1999 110 57 S/GA LTE WIRE ROPE STS 8245 615 1690 FTI Oct 31 1999 111 54 S/G A LTE WIRE ROPE STB 8245 610 1690 FTI Oct 31 1999 112 66 S/G A LTE WIRE ROPE CF36-2 STB 18 1690 FTI Oct 31 1999 113 66 S/G A LTE WIRE ROPE 8245 608 1690 FTI Oct 31 1999 STB 4 41 S/GB UTE FLEX 398-730 STB 131 1600 BW May 1 1992 25 9 S/G B UTE FLEX 398 STB 137 1600 BW May 1 1992 25 9 S/G B LTE FLEX 398-730 STB 132 1600 BW May 1 1992 26 10 S/G B LTE WIRE ROPE 9654-1 STB 188 1690 6W May 81994 26 10 S/G B UTE WIRE ROPE 9654-1 STB 177 1690 aW May 81994 29 75 S/G B LTE FLEX 398-730 STB 40 1600 BW May 1 1992 31 37 S/G B LTE WIRE ROPE 9654-1 STB 191 1690 FTI Jul 22 1997 34 61 S/G B UTE FLEX NX5398 STB 140 1600 8W Mar 1 1990 35 38 S/G B LTE WIRE ROPE 9654-1 STB 190 1690 FTI Jul 22 1997 39 115 S/G B UTE WIRE ROPE 9654-1 STB 158 1690 FTI Jul 21 1997 41 41 S/G B LTE FLEX 398-730 STB 68 1600 BW May 1 1992 44 45 S/G B LTE FLEX 398-730 51 STB 1600 BW May 1 1992 44 46 S/G B LTE WIRE ROPE 9654-1 STB 187 1690 FTI Jul 22 1997 44 49 S/G B LTE FLEX 398-730 110 1600 BW STB May 1 1992 45 50 S/G B LTE FLEX 398-730 125 ST8 1600 8W May 1 1992 46 1 S/GB UTE FLEX NX5398 127 STB 1600 BW Mar 1 1990 46 83 S/G B LTE WIRE ROPE 9654-1 150 BW ST8 Mar 19 1996 47 2 S/G B UTE FLEX NX5398 119 1600 BW STB Mar 1 1990 48 1 S/GB UTE FLEX NX5398 27 STB 1600 8W Mar 1 1990 48 2 S/G B" UTE FLEX NX5398 132 STB 1600 BW Mar 1- 1990 50 35 S/G 8 LTE WIRE ROPE 9654-1 STB 178 BW Apr 2 1996 50 38 S/G B LTE FLEX 398-730 92 1600 sW May 1 1992 Page 15 of 20

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RJFRAMATOME ANP 51-5000475-01 51 39 S/G B LTE STB FLEX 398-730 121 1600 BW May 1 1992 51 81 S/G B LTE STB FLEX 398-730 104 1600 BW May 1 1992 53 40 S/G B LTE STB FLEX 398-730 27 1600 BW May 1 1992 54 41 S/G B LTE STB FLEX 398-730 45 1600 BW May 1 1992 56 28 S/G B LTE STB WIRE ROPE 9654-1 170 BW Mar19 1996 57 38 S/G B LTE STB WIRE ROPE 9654-1 189 1690 FTI Jul 22 1997 58 38 SIG B LTE STB WIRE ROPE 9654-1 167 BW Apr 2 1996 59 32 S/G B LTE STB FLEX 398-730 97 1600 BW May 1 1992 62 125 S/G B UTE STB FLEX 398-730 48 1600 8W May 1 1992 63 30 SIG B LTE STB FLEX 398-730 124 1600 BW May 1 1992 68 6 SIG B UTE STB WIRE ROPE 9654-1 166 1690 FTI Jul21 1997 69 5 SiG B UTE STB WIRE ROPE 9654-1 162 1690 FTI Jul 21 1997 70 1 SIG B SLV ROLLED 764371 502 1690 BW Apr 2 1994 70 5 S/G B UTE STS WIRE ROPE 9654-1 157 1690 FTI Jul21 1997 71 1 S/G B SLV ROLLED 764371 474 1690 BW Apr 21994 71 2 S/G B SLV ROLLED 764371 492 1690 BW Apr 21994 71 3 S/G B SLV ROLLED 764371 723 1690 BW Apr 21994 72 1 SiG B SLV ROLLED 764371 146 1690 BW Apr 2 1994 72 2 S/G B SLV ROLLED 764371 149 1690 BW Apr 2 1994 72 3 S/G B SLV ROLLED 764371 463 1690 BW Apr 2 1994 72 4 SIG B SLV ROLLED 764371 137 1690 BW Apr 2 1994 72 5 S/G B UTE STB WIRE ROPE 9654-1 181 1690 FTI Jul 21 1997 72 55 S/G B UTE STB WIRE ROPE 9654-1 160 1690 FTI Jul21 1997 73 1 S/G B UTE STB FLEX 398-730 132 1600 BW May 1 1992 73 2 S/G B SLV ROLLED 764371 745 1690 BW Apr 2 1994 73 3 SIG B SLV ROLLED 764371 150 1690 BW Apr 2 1994 73 4 SIG B SLV ROLLED 764371 136 1690 BW Apr 21994 73 5 SIG B SLV ROLLED 764371 217 1690 8W Apr 2 1994 73 6 SIG B SLV ROLLED 764371 178 1690 8W Apr 2 1994 73 19 SIG B UTE STS FLEX 398-730 107 1600 8W May 1 1992 73 24 S/G B UTE STB WIRE ROPE 9654-01 165 BW Mar18 1996 73 35 SIG B UTE STB WIRE ROPE 9654-1 168 1690 FTI Jul 21 1997 73 37 SfG B UTE STB WIRE ROPE 9654-1 174 1690 FTI Jul21 1997 73 39 S/G B LTE STB WIRE ROPE 9654-1 186 1690 FTI Jul 22 1997 73 66 S/G B UTE STB WIRE ROPE 9654-1 151 1690 FTI Jul21 1997 74 1 SIG B SLV ROLLED 764371 295 1690 BW Apr 2 1994 74 2 S/G B SLV ROLLED 764371 607 1690 BW Apr 2 1994 74 3 SIG B SLV ROLLED 764371 617 1690 BW Apr 2 1994 74 4 S/G B SLV ROLLED 764371 571 1690 8W Apr 2 1994 74 5 SIG B SLV ROLLED 764371 586 1690 8W Apr 21994 74 6 SIG B SLV ROLLED 764371 209 1690 BW Apr 21994 74 7 S0G B SLV ROLLED 764371 160 1690 8W Apr 21994 74 8 SIG B SLV ROLLED 764371 158 1690 BW Apr 2 1994 74 9 SIG B SLV ROLLED 764371 177 1690 BW Apr 21994 74 10 SIG B SLV ROLLED 764371 226 1690 8W Apr 2 1994 74 11 SIG B SLV ROLLED 764371 185 1690 BW Apr 21994 74 12 S/G B SLV ROLLED 764371 139 1690 BW Apr 2 1994 74 13 S/G B "SLV ROLLED 764371 214 1690 BW Apr 2 1994 74 14 S/G B SLV ROLLED 753957 1086 1690 BW Apr 2 1994 74 15 SIG B SLV ROLLED 753957 1097 1690 BW Apr 2 1994 74 16 S/G B UTE STB FLEX 398-730 126 1600 BW May 1 1992 74 17 SfG B SLV ROLLED 753957 1085 1690 BW Apr 2 1994 74 18 S/G B SLV ROLLED 753957 1103 1690 BW Apr 2 1994 74 19 S/G B SLV ROLLED 753957 1067 1690 BW Apr 2 1994 74 20 S/G B SLV ROLLED 753957 11061690 8W Apr 2 1994 74 21 S/G B .SLV ROLLED 753957 1071 1690 8W Apr 2 1994 74 22 S/G B UTE STB FLEX 398-730 73 1600 BW May 1 1992 74 23 SIG B SLV ROLLED 753957 11141690 BW Apr 2 1994 Page 16 of 20

/FRAMATOME ANP 51-5000475-01 74 24 S/G B SLV ROLLED 753957 1102 1690 BW Apr 2 1994 74 25 S/G B SLV ROLLED 753957 1096 1690 8W Apr 21994 74 26 S/G B SLV ROLLED 753957 1101 1690 BW Apr 21994 74 27 S/G B SLV ROLLED 753957 1100 1690 BW Apr 21994 74 28 S/G B SLV ROLLED 753957 1081 1690 BW Apr 2 1994 74 29 SIG B SLV ROLLED 764371 199 1690 BW Apr 2 1994 74 31 S/ GB UTE STB WIRE ROPE 965' 4-1 179 1690 FTI Jul 21 1997 74 34 S/G B UTE STB WIRE ROPE 96544-1 172 1690 FT1 Jul 21 1997 74 48 S/G B UTE STB WIRE ROPE 96544-1 163 1690 FTI Jul 21 1997 74 49 S/G B UTE STB WIRE ROPE 96544-1 180 1690 FTI Jul 21 1997 75 1 S/G B SLV ROLLED 764371 606 1690 8W Apr 21994 75 2 S/G B SLV ROLLED 764371 603 1690 BW Apr 2 1994 75 3 S/G 8 "SLV ROLLED 764371 498 1690 BW Apr 2 1994 75 4 SIG B S LV ROLLED 764371 614 1690 BW Apr 2 1994 75 5 SIG B SLV ROLLED 764371 305 1690 BW Apr 2 1994 75 6 S/G B SLV ROLLED 764371 589 1690 BW Apr 2 1994 75 7 S/G B SLV ROLLED 764371 584 1690 8W Apr 2 1994 75 8 S/G B SLV ROLLED 764371 619 1690 BW Apr 2 1994 75 9 SIG B SLV ROLLED 764371 591 1690 BW Apr 2 1994 75 10 S/G B SLV ROLLED 764371 329 1690 BW Apr 2 1994 75 11 S/G B SLV ROLLED 764371 393 1690 8W Apr 2 1994 75 12 S/G B UTE STB SGMT  ? 1600 BW Sep 1 1979

?

75 13 S/G B UTE STB SGMT  ? 1600 BW Sep 1 1979 75 14 S/G B SLV ROLLED 764371 503 1690 BW Apr 2 1994 75 15 SIG B SLV ROLLED 764371 602 1690 BW Apr 2 1994 75 16 S/G B SLV ROLLED 764371 314 1690 BW Apr 2 1994 75 17 S/G B UTE STB SGMT  ? 1600 BW Sep 1 1979 75 18 S/G B UTE STB SGMT  ? 1600 BW Sep 1 1979 75 19 S/G 8 SLV ROLLED 764371 325 1690 8W Apr 2 1994 75 20 S/G B UTE STB SGMT  ? 1600 8W Sep 1 1979 75 21 S/G B SLV ROLLED 764371 337 1690 BW Apr 21994 75 22 S/G B SLV ROLLED 764371 375 1690 BW Apr 21994 75 23 S/G B SLV ROLLED 764371 344 1690 8W Apr 2 1994 75 24 SIG B SLV ROLLED 764371 282 1690 BW Apr 21994 75 25 S/G B SLV ROLLED 764371 311 1690 BW Apr 21994 75 26 S/G B SLV ROLLED 764371 272 1690 BW Apr 21994 75 27 S/G B "SLV ROLLED 764371 273 1690 BW Apr 2 1994 75 28 S/G B SLV ROLLED 764371 315 1690 BW Apr 2 1994 75 29 SIG B SLV ROLLED 764371 345 1690 BW Apr 2 1994 75 30 S/G B SLV ROLLED 764371 276 1690 BW Apr 2 1994 75 31 S/G B SLV ROLLED 764371 322 1690 BW Apr 2 1994 75 32 S/G B SLV ROLLED 764371 383 1690 8W Apr 2 1994 75 33 S/G B SLV ROLLED 764371 297 1690 BW Apr 2 1994 75 34 S/G B SLV ROLLED 764371 364 1690 BW Apr 2 1994 75 35 S/G B SLV ROLLED 764371 398 1690 BW Apr 2 1994 75 36 S/G B SLV ROLLED 764371 319 1690 BW Apr 2 1994 75 37 S/G B SLV ROLLED 764371 403 1690 BW Apr 21994 75 38 S/G B SLV ROLLED 764371 288 1690 BW Apr 2 1994 75 39 S/G B SLV ROLLED 764371 367 1690 BW Apr 2 1994 75 40 S/G B SLV ROLLED 764371 333 1690 BW Apr 2 1994 75 41 S/G B SLV ROLLED 764371 371 1690 BW Apr 2 1994 75 42 S/G B SLV ROLLED 764371 356 1690 BW Apr 2 1994 75 43 S/G B SLV ROLLED 764371 300 1690 BW Apr 2 1994 75 44 S/G B SLV ROLLED 764371 321 1690 BW Apr 2 1994 75 45 S/G 8 UTE STB SGMT  ? 1600 BW Sep 1 1979 75 46 S/G 8 SLV ROLLED 764371 290 1690 BW Apr 2 1994 75 47 SIG B SLV ROLLED 764371 316 1690 BW Apr 2 1994 75 55 S/G B UTE STB SGMT  ?  ? 1600 BW Sep 1 1979 Page 17 of 20

fFRAMATOME ANP 51-5000475-01 75 63 SIG B UTE STB SGMT  ?  ? 1600 BW Sep 1 1979 77 1 S/G B SLV ROLLED 764371 324 1690 BW Apr 2 1994 77 2 S/G B SLV ROLLED 764371 281 1690 BW Apr 2 1994 77 3 S/G B SLV ROLLED 764371 335 1690 BW Apr 2 1994 77 4 SIG B SLV ROLLED 764371 334 1690 BW Apr 2 1994 77 5 S/G B SLV ROLLED 764371 287 1690 BW Apr 2 1994 77 6 S/G B SLV ROLLED 764371 312 1690 8W Apr 2 1994 77 7 S/G 8 SLV ROLLED 764371 336 1690 8W Apr 2 1994 77 8 S/G B SLV ROLLED 764371 289 1690 8W Apr 2 1994 77 9 S/G B SLV ROLLED 764371 275 1690 BW Apr 2 1994 77 10 S/G B SLV ROLLED 764371 320 1690 BW Apr 2 1994 77 11 S/G B SLV ROLLED 764371 292 1690 8W Apr 21994 77 12 S/G B UTE STS FLEX 398-730 136 1600 BW May 1 1992 77 13 S/G B SLV ROLLED 764371 318 1690 BW Apr 2 1994 77 14 S/G B SLV ROLLED 764371 386 1690 8W Apr 2 1994 77 15 S/G 8 SLV ROLLED 764371 327 1690 BW Apr 2 1994 77 16 S/G B SLV ROLLED 764371 285 1690 BW Apr 2 1994 77 17 S/G B SLV ROLLED 764371 307 1690 8W Apr 2 1994 77 18 S/G B SLV ROLLED 764371 277 1690 BW Apr 2 1994 77 19 S/G B SLV ROLLED 764371 323 1690 BW Apr 2 1994 77 20 S/G B SLV ROLLED 764371 381 1690 BW Apr 2 1994 77 21 S/G B SLV ROLLED 764371 369 1690 BW Apr 2 1994 77 22 S/G B SLV ROLLED 764371 354 1690 BW Apr 2 1994 77 23 S/G B SLV ROLLED 764371 331 1690 BW Apr 2 1994 77 24 S/G B SLV ROLLED 764371 283 1690 BW Apr 2 1994 77 25 S/GB UTE STB SGMT  ? 1600 8W Sep 1 1979 77 26 S/G B SLV ROLLED 764371 298 1690 BW Apr 2 1994 77 27 SIG B SLV ROLLED 764371 358 1690 BW Apr 2 1994 77 28 S/G B SLV ROLLED 764371 274 1690 8W Apr 2 1994 77 29 S/G B SLV ROLLED 764371 271 1690 BW Apr 2 1994 77 30 SlG 8 SLV ROLLED 764371 302 1690 BW Apr 2 1994 77 31 S/G B SLV ROLLED 764371 278 1690 BW Apr 2 1994 77 32 S/G B SLV ROLLED 764371 293 1690 BW Apr 2 1994 77 33 S/G 8 SLV ROLLED 764371 328 1690 BW Apr 2 1994 77 34 S/G B UTE STB SGMT  ?  ? 1600 8W Sep 1 1979 77 35 S/G B SLV ROLLED 764371 279 1690 BW Apr 2 1994 77 36 S/G B SLV ROLLED 764371 304 1690 BW Apr 2 1994 77 37 S/G B SLV ROLLED 764371 286 1690 8W Apr 2 1994 77 38 S/G B SLV ROLLED 764371 306 1690 BW Apr 2 1994 77 39 SIG B SLV ROLLED 764371 332 1690 BW Apr 2 1994 77 40 S/G B SLV ROLLED 764371 299 1690 BW Apr 2 1994 77 42 S/G B UTE STB SGMT  ?  ? 1600 BW Sep 1 1979 77 43 S/G B SLV ROLLED 764371 326 1690 BW Apr 2 1994 77 44 S/G B SLV ROLLED 764371 294 1690 BW Apr 2 1994 77 45 SIG B SLV ROLLED 764371 310 1690 BW Apr 2 1994 77 46 S/G B SLV ROLLED 764371 291 1690 BW Apr 2 1994 77 47 S/G B SLV ROLLED 764371 280 1690 8W Apr 2 1994 77 48 S/G B UTE STB WIRE ROPE 9654-01 152 8W Mar 18 1996 77 49 S/G B UTE STB SGMT  ?  ? 1600 BW Sep 1 1979 77 50 SIG B UTE STB SGMT  ?  ? 1600 8W Sep 1 1979 77 60 S/G B UTE STB SGMT  ?  ? 1600 8W Sep 1 1979 78 1 S/G B SLV ROLLED 764371 766 1690 BW Apr 2 1994 78 2 S/G B SLV ROLLED 764371 477 1690 BW Apr 2 1994 78 3 S/G B SLV ROLLED 764371 730 1690 BW Apr 2 1994 78 4 S/G 8 SLV ROLLED 764371 582 1690 BW Apr 2 1994 78 5 S/G 8 SLV ROLLED 764371 600 1690 8W Apr 2 1994 78 6 S/G B SLV ROLLED 764371 596 1690 BW Apr 2 1994 78 7 S/G B SLV ROLLED 764371 724 1690 8W Apr 2 1994 Page 18 of 20

FRAMATOME AN P 51-5000475-01 78 8 SfG B SLV ROLLED 764371 618 1690 BW Apr 2 1994 78 9 S/G B SLV ROLLED 764371 594 1690 BW Apr 2 1994 78 10 SfG B SLV ROLLED 764371 587 1690 8W Apr 2 1994 78 11 S/G B SLV ROLLED 764371 705 1690 BW Apr 2 1994 78 12 S/G B SLV ROLLED 764371 575 1690 BW Apr 2 1994 78 13 SfG B UTE STB FLEX 398-730 36 1600 BW May 1 1992 78 14 SfG B SLV ROLLED 764371 576 1690 BW Apr 2 1994 78 15 S/G B SLV ROLLED 764371 612 1690 BW Apr 2 1994 78 16 SfG B SLV ROLLED 764371 590 1690 6W Apr 2 1994 78 17 SfG B SLV ROLLED 764371 494 1690 BW Apr 2 1994 78 18 S/G 8 SLV ROLLED 764371 712 1690 BW Apr 2 1994 78 19 SfG B SLV ROLLED 764371 748 1690 BW Apr 2 1994 78 20 S/G B SLV ROLLED 764371 734 1690 6W Apr 2 1994 78 21 SfG B SLV ROLLED 764371 763 1690 8W Apr 2 1994 78 22 SfG B SLV ROLLED 764371 744 1690 BW Apr 2 1994 78 23 SfG B SLV ROLLED 764371 729 1690 8W Apr 2 1994 78 24 S/G B SLV ROLLED 764371 714 1690 8W Apr 2 1994 78 25 SfG B SLV ROLLED 764371 726 1690 BW Apr 2 1994 78 26 SfG B SLV ROLLED 764371 722 1690 BW Apr 2 1994 78 27 S/G B SLV ROLLED 764371 702 1690 BW Apr 2 1994 78 28 S/G B SLV ROLLED 764371 704 1690 6W Apr 2 1994 78 29 S/G B SLV ROLLED 764371 725 1690 BW Apr 2 1994 78 49 SIG B UTE STB FLEX NX5398 133 1600 BW Mar 1 1990 78 51 S/G B UTE STB FLEX 398-730 82 1600 BW May 1 1992 79 1 SfG B SLV ROLLED 764371 728 1690 6W Apr 2 1994 79 2 S/G B SLV ROLLED 764371 770 1690 BW Apr 2 1994 79 3 SfG B SLV ROLLED 764371 727 1690 BW Apr 2 1994 79 4 S/G B UTE STB FLEX 398-730 114 1600 8W May 1 1992 79 5 S/G B SLV ROLLED 764371 794 1690 BW Apr 2 1994 79 6 SfG B SLV ROLLED - 764371 680 1690 BW Apr 2 1994 79 65 SfG B UTE STB WIRE ROPE 9654-1 164 1690 FTI Jul 21 1997 80 1 S/G B SLV ROLLED 764371 743 1690 BW Apr 2 1994 80 2 SfG B SLV ROLLED 764371 507 1690 6W Apr 2 1994 80 3 SfG B SLV ROLLED 764371 768 1690 BW Apr 2 1994 80 4 S/G 8 SLV ROLLED 764371 706 1690 6W Apr 2 1994 81 1 SIG B SLV ROLLED 764371 267 1690 6W Apr 2 1994 81 2 SfG B SLV ROLLED 764371 475 1690 BW Apr 2 1994 81 3 S/G B SLV ROLLED 764371 750 1690 BW Apr 2 1994 81 95 SfG 6 LTE STB FLEX 398-730 38 1600 BW May 1 1992 84 3 SfG B UTE STB WIRE ROPE 9654-1 175 [690 FTI Jul 21 1997 84 35 SfG B LTE STB FLEX 398-730 52 1600 BW May 1 1992 86 31 SfG B LTE STB FLEX 398-730 67 1600 BW May 1 1992 89 34 SIG B LTE STB WIRE ROPE 9654-1 161 BW Apr 6 1996 92 28 S/G B LTE STB WIRE ROPE 9654-1 154 1690 BW May 8 1994 92 29 SIG B LTE STB FLEX 398-730 54 1600 BW May 1 1992 93 37 S0G B LTE STB FLEX 398-730 111 1600 BW May 1 1992 94 93 S/G 8 LTE STB FLEX 398-730 35 1600 8W May 1 1992 95 43 S/G B LTE STB FLEX 398-730 133 1600 6W May 1 1992 96 33 SfG B LTE STB FLEX NX3624- 226 1600 BW Mar 1 1990 96 92 SfG B LTE STB FLEX 398-730 137 1600 BW May 1 1992 96 116 SfG B UTE STB WIRE ROPE 9654-1 185 BW Apr 3 1996 97 43 SIG B LTE STB WIRE ROPE 9654-1 176 1690 6W May 8 1994 98 28 SfG B LTE STB FLEX NX3624- 241 1600 BW Mar 1 1990 98 44 SfG B LTE STB FLEX NX4064G 629 1600 BW Mar 1 1990 99 91 S/G B LTE STB FLEX 398-730 119 1600 6W May 1 1992 102 37 SfG B LTE STB FLEX 398-730 134 1600 8W May 1 1992 103 43 S/G B LTE STS FLEX 398-730 123 1600 BW May 1 1992 105 33 SfG B LTE STB FLEX 398-730 77 1600 6W May 1 1992 Page 19 of 20

SFRAMATOME ANP 51-5000475-01 107 35 S/G B LTE STB FLEX 398-730 106 1600 BW May 1 1992 108 45 S/G B LTE STB WIRE ROPE CF36-2 17 1690 FTI Jul 23 1997 109 34 S/G B LTE STB FLEX NX3624 237 1600 BW Mar 1 1990 112 39 S/G B LTE STB FLEX 398-730 70 8600 BW May 1 1992 116 46 S/G B LTE STB FLEX NX3624 234 1600 BW Mar 1 1990 122 21 S/G B LTE STB WIRE ROPE 9654-1 155 BW Mar 19 1996 125 9 S/G B UTE STB FLEX NX5398 15 1600 BW Mar 1 1990 125 9 S/G B LTE STB FLEX NX3624 236 1600 BW Mar 1 1990 133 33 SIG B LTE STB SGMT NX3232G 506 1600 BW May 1 .1992 133 33 S/G B LTE STB SGMT NX3232G 507 1600 BW May 1 1992 133 33 SIG B LTE STB SGMT NX3232G 511 1600 BW May 1 1992 133 33 SIG B LTE STB SGMT NX3232G 517 1600 BW May 1 1992 133 33 S/G B LTE STB SGMT NX3232G 548 1600 BW May 1 1992 133 33 S/G B LTE STB SGMT NX3232G 557 1600 BW May 1 1992 133 33 S/G B LTE STB SGMT NX3232G 559 1600 BW May 1 1992 133 33 S/G B LTE STB SGMT NX3232G 564 1600 BW May 1 1992 133 33 S/G B LTE STB SGMT NX3232G 567 1600 BW May 1 1992 133 33 SfG B LTE STB SGMT NX3232G 577 1600 8W May 1 1992 133 33 SiG B LTE ST8 SGMT NX3232G 586 1600 BW May 1 1992 133 33 SIG B LTE STB SGMT NX3232G 593 1600 BW May 1 1992 133 33 SIG 8 LTE STB SGMT NX3232G 595 1600 BW May 1 1992 133 33 SIG B LTE STB SGMT NX3232G 600 1600 BW May 1 1992 133 33 SfG B LTE STB SGMT NX3232G 602 1600 BW May 1 1992 138 75 S/G B UTE STB FLEX 398-730 122 1600 BW May 1 1992 71 131 S/G B UTE STB WIRE ROPE 8245 154 1690 FTI Oct 29 1999 Total Indications Found = 539 Note: Some of the data fields listed in this appendix contain "?". These fields were not entered into the Framatome ANP FDMS database and therefore are identified as "?" in this appendix. Omission of this data is not pertinent to the results or conclusions of this evaluation.

Page 20 of 20

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ATTACHMENT C LICENSE AMENDMENT REQUEST #263, REVISION 0 Relocation Of Reactor Coolant System Parameters To The Core Operating Limits Report And 20 Percent Steam Generator Tube Plugging Framatome ANP Affidavit of Proprietary Information

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is James F. Mallay. I am Director, Regulatory Affairs, for Framatome ANP ("FRA-ANP"), and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by FRA-ANP to determine whether certain FRA-ANP information is proprietary. I am familiar with the policies established by FRA-ANP to ensure the proper application of these criteria.

3. I am familiar with the information prepared by FRA-ANP and contained in a calculational file, "Flow-Induced Vibration Analysis of TMI OTSG Tubes Due to Power Up-Rate,"

32-1257514-01. This file is referred to herein as "Document." Information contained in this Document has been classified by FRA-ANP as proprietary in accordance with the policies established by FRA-ANP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FRA-ANP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

6. The following criteria are customarily applied by FRA-ANP to determine whether information should be classified as proprietary:

(a) The information reveals details of FRA-ANP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FRA-ANP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for FRA-ANP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by FRA-ANP, would be helpful to competitors to FRA-ANP, and would likely cause substantial harm to the competitive position of FRA-ANP.

7. In accordance with FRA-ANP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside FRA-ANP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. FRA-ANP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

3 SUBSCRIBED before me this day of 2002.

ia , ,Z-ý Ella F. Carr-Payne NOTARY PUBLIC, STATE OF VIRGINIA MY COMMISSION EXPIRES: 8/31/05