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Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 (Exelon Generation Company, LLC (Exelon) - Section 4 -Time-limited Aging Analys
	ML042050503
Person / Time
Site:	Dresden, Quad Cities
Issue date:	07/01/2004
From:	Kuo P; Division of Regulatory Improvement Programs
To:	Skolds J; Exelon Generation Co
	Burton W, NRR/DRIP/RLEP, 301-415-2853
Shared Package
ML042050507	List: ML042050493; ML042050495; ML042050500; ML042050503; ML042050505; ML042050508;
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4. TIME-LIMITED AGING ANALYSES 4.1 Identification of Time-Limited Aging Analyses This section addresses the identification of time-limited aging analyses (TLAAs). The applicant discusses the TLAAs in Sections 4.2 through 4.7 of the license renewal application (LRA). The staffs review of the TLAAs can be found in Sections 4.2 through 4.7 of this safety evaluation report (SER).

The TLAAs are plant-specific safety analyses that are based on an explicitly assumed 40-year plant life. Pursuant to Title 10, Section 54.21(c)(1), of the Code of Federal Regulations (10 CFR 54.21(c)(1)), the applicant for license renewal must provide a list of TLAAs, as defined in 10 CFR 54.3.

In addition, pursuant to 10 CFR 54.21(c)(2), an applicant must provide a list of plant-specific exemptions granted under 10 CFR 50.12 that are based on TLAAs. For any such exemptions, the applicant must provide an evaluation that justifies the continuation of the exemptions for the period of extended operation.

4.1.1 Summary of Technical Information in the Application A list of potential generic TLAAs was assembled using the scope and methods for identifying TLAAs consistent with NUREG-1800, Nuclear Energy Institute (NEI) 95-10, the Statements of Consideration for 10 CFR Part 54, and prior license renewal applications. The current licensing basis (CLB) for the Dresden/Quad Cities Nuclear Power Station (D/QCNPS), including the updated final safety analysis report (UFSAR) and design-basis documents, was searched to identify plant-specific and unit-specific TLAAs. The resulting list of potential TLAAs was screened against the six 10 CFR 54.3(a) criteria and divided into the following general TLAA categories:

neutron embrittlement of the reactor vessel and internals metal fatigue of the reactor vessel, internals, and primary coolant boundary piping and components environmental qualification of electrical equipment loss of prestress concrete containment tendons fatigue of the primary containment, attached piping, and components other plant-specific TLAAs Information about the TLAAs in a category is described as to applicability, summary description, analysis, and disposition in accordance with 10 CFR 54.21(c)(1).

The applicant searched docketed correspondence, the operating licenses, and the UFSARs to identify any exemptions in effect, pursuant to 10 CFR 50.12. The applicant stated that the

4-2 identified exemptions were evaluated to determine if they involved TLAAs as defined in 10 CFR 54.3. No exemptions based on a TLAA as defined in 10 CFR 54.3 were identified.

4.1.2 Staff Evaluation In LRA Section 4.1, the applicant identified the TLAAs applicable to Dresden and Quad Cities and discussed exemptions based on TLAAs. The staff reviewed the information to determine whether the applicant provided adequate information to meet the requirements of 10 CFR 54.21(c)(1) and 10 CFR 54.21(c)(2).

As indicated by the applicant, TLAAs are defined in 10 CFR 54.3 as analyses that meet the following six criteria:

(1) involve systems, structures, and components within the scope of license renewal, as delineated in Section 54.4(a)

(2) consider the effects of aging (3) involve time-limited assumptions defined by the current operating term (for example, 40 years)

(4) were determined to be relevant by the licensee in making a safety determination (5) involve conclusions or provide the basis for conclusions related to the capability of the system, structure, and component to perform its intended functions, as delineated in Section 54.4(b)

(6) are contained or incorporated by reference in the CLB The applicant listed the TLAAs applicable to Dresden and Quad Cities, both jointly and individually, in Table 4.1-1 of the LRA. The staff reviewed the categorization of the TLAAs for conformance with Tables 4.1-2 and 4.1-3 in NUREG-1800 and potential TLAAs that were identified from the review of other license renewal applications.

4.1.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable list of TLAAs as required by 10 CFR 54.21(c)(1), and has confirmed that no 10 CFR 50.12 exemptions have been granted on the basis of a TLAA, as required by 10 CFR 54.21(c)(2).

4.2 Reactor Vessel and Internals Neutron Embrittlement During plant service, neutron irradiation reduces the fracture toughness of ferritic steel in the reactor vessel beltline region of light-water nuclear power reactors. Areas of review to ensure that the reactor vessel has adequate fracture toughness to prevent brittle failure during normal and off-normal operating conditions are (1) upper-shelf energy (USE), (2) adjusted reference temperature (ART), (3) a low-pressure coolant injection (LPCI) reflood thermal shock analysis, (4) heatup and cooldown (pressure-temperature limit) curves, and (5) Boiling Water Reactor

4-3 (BWR) Vessel and Internals Project (VIP) VIP-05 analysis for elimination of circumferential weld inspection, and (6) analysis of the axial welds. The adequacy of the analyses for these six areas is reviewed for the period of extended operation.

The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT),

the mean value of the adjustment in reference temperature caused by irradiation (delta RTNDT),

and a margin (m) term. The delta RTNDT is the product of a chemistry factor (CF) and a fluence factor. The chemistry factor is dependent upon the amount of copper and nickel in the material and may be determined from tables in Regulatory Guide (RG) 1.99, Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the fluence, and the calculation methods. Revision 2 of RG 1.99 describes the methodology to be used in calculating the margin term. The mean RTNDT is the sum of the initial RTNDT and the delta RTNDT, without the margin term.

The ART values are used in the analysis for the adjusted reference temperature for the reactor vessel material because of neutron embrittlement, the pressure-temperature limits, and the reflood thermal shock. The mean RTNDT values are used in the analysis of the circumferential weld examination relief and the axial weld failure probability.

4.2.1 Summary of Technical Information in the Application The applicant described its evaluation of this TLAA in LRA Section 4.2, Neutron Embrittlement of the Reactor Vessel and Internals. In order to demonstrate that neutron embrittlement does not significantly impact BWR reactor pressure vessel (RPV) and vessel internals integrity during the license renewal term, the applicant included discussion of the following topics related to neutron embrittlement in LRA Section 4.2:

reactor vessel materials upper-shelf energy reduction due to neutron embrittlement (LRA Section 4.2.1) adjusted reference temperature for reactor vessel materials due to neutron embrittlement (LRA Section 4.2.2) reflood thermal shock analysis of the reactor vessel (LRA Section 4.2.3) reflood thermal shock analysis of the reactor vessel core shroud and repair hardware (LRA Section 4.2.4) reactor vessel thermal limit analysesoperating pressure-temperature limits (LRA Section 4.2.5) reactor vessel circumferential weld examination relief (LRA Section 4.2.6) reactor vessel axial weld failure probability (LRA Section 4.2.7)

4-4 4.2.1.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement Appendix G of 10 CFR Part 50 requires that the predicted end-of-life Charpy USE for reactor vessel materials be at least 50 ft-lb, unless an approved analysis supports a lower value. The applicant determined the 54 effective full-power year (EFPY) fluence for the Dresden and Quad Cities reactor vessels using the methodology of NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation, which was approved by the U.S.

Nuclear Regulatory Commission (NRC) in a letter dated September 14, 2001, from Mr. S.A.

Richards (NRC) to Mr. J.F. Klapproth (General Electric (GE)). The applicant performed one bounding fluence calculation for D/QCNPS and determined peak fluence at the vessel inner surface and at the 1/4 vessel thickness location (1/4T) for evaluating USE. Initial unirradiated test data are not available for the D/QCNPS reactor vessels to demonstrate a minimum 50 ft-lb USE using the standard methods. Therefore, the applicant evaluated the 54-EFPY USE by an equivalent margin analysis methodology approved by the NRC in NEDO-32205-A. The applicant used calculated 54-EFPY fluence and D/QCNPS surveillance capsule results for evaluating the 54-EFPY USE. The results are presented in LRA Tables 4.2.1-1 through 4.2.1-

8. The results show that the percent reductions in USE for limiting beltline plates and welds for all four D/QCNPS units are less than the BWRVIP-74 equivalent margin analysis acceptance criteria. The applicant stated that a report summarizing the results of the equivalent margin analysis will be submitted for NRC approval by December 31, 2003. The applicant further stated that the 54-EFPY USE values will be managed in conjunction with the surveillance capsule results from the BWRVIP integrated surveillance program.

4.2.1.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement Neutron irradiation causes an increase in the ART of the RPV beltline materials. Tables 4.2.2-1 and 4.2.2-2 of the LRA provide the 54-EFPY peak fluence, shift in initial nil-ductility transition reference temperature (delta RTNDT), and ART, respectively, for Dresden and Quad Cities units.

In these tables, the applicant provided the delta RTNDT and ART values only for one limiting material. The applicant stated that because of the refinement in the approved methodology used to calculate the 54-EFPY fluence, the material with the limiting ART is the axial weld, with the exception of Dresden Unit 3 where the axial weld and girth weld ART values are identical.

The applicant further stated that the use of American Society of Mechanical Engineers (ASME)

Code Case N-588 for Dresden Unit 3 causes the axial weld to become the limiting material.

Therefore, data for a single limiting material were presented in LRA Tables 4.2.2-1 and 4.2.2-2.

4.2.1.3 Reflood Thermal Shock Analysis of the Reactor Vessel In LRA, Section 4.2.3, the applicant stated the following:

The Dresden and Quad Cities UFSARs describe an end-of-life thermal shock analysis performed on the reactor vessels for a design basis LOCA followed by a low-pressure coolant injection. The effects of embrittlement assumed by this thermal shock analysis will change with an increase in operating period. This analysis satisfies the criteria of 10 CFR 54.3(a). As such, this analysis is a TLAA.

Thermal shock analysis of the RPV considers a design-basis loss-of-coolant accident (LOCA) followed by an LPCI accounting for the effects of neutron embrittlement at the end-of-life (54

4-5 EFPYs). The original analysis has been superseded by an analysis for BWR-6 vessels that is applicable to the D/QCNPS BWR-3 reactor vessels. The revised analysis assumes end-of-life material toughness, which in turn depends on end-of-life ART. The critical location for fracture mechanics analysis is at the 1/4T location. For this event, the peak stress intensity occurs at approximate 300 seconds after the LOCA. At that time, the temperature at 1/4T is approximately 204 C (400 F), which is much higher than the 54-EFPY ART of 40 C (104

F) for the limiting material of the D/QCNPS vessels. Therefore, the applicant indicated that the revised analysis is valid for the period of extended operation.

4.2.1.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware In LRA, Section 4.2.4, the applicant stated the following:

Radiation embrittlement may affect the ability of reactor vessel internals, particularly the core shroud and repair hardware, to withstand a low-pressure coolant injection (LPCI) thermal shock transient. Core shroud repair hardware was installed on the Dresden and Quad Cities core shrouds after 20 years of operation when cracks were found on the shroud. The analysis of core shroud strain due to reflood thermal shock is a TLAA because it is part of the current license basis, supports a safety determination, and is based on the calculated lifetime neutron fluence.

The thermal shock analysis of D/QCNPS reactor vessel core shrouds considers the embrittlement effects of 54-EFPY fluence at the inside surface opposite to the midpoint of the fuel centerline where the core shroud receives the maximum irradiation. The applicant calculated 54-EFPY fluence for that location on the reactor vessel core shroud using the methodology of NEDC-32983P, which was approved by the NRC. The calculated 54-EFPY fluence at that location was 5.85 x 1020 n/cm2. The applicant stated that the calculated thermal strain amplitude at the most irradiated location is 0.3 percent. The applicant further stated that according to the GE document Y1002A602, Revision 3, 304 Stainless Steel, Irradiated, October 16, 1985, the allowable value of the thermal strain for irradiation levels in excess of 1 x 1021 n/cm2 and for this faulted event is at least 20 percent, which bounds the 0.3 percent thermal strain amplitude in Dresden and Quad Cities. Therefore, the applicant indicated that the peak thermal shock strain location is acceptable considering the embrittlement effects of a 60-year (54-EFPY) operating period.

4.2.1.5 Reactor Vessel Thermal Limit AnalysesOperating Pressure-Temperature Limits The applicant used the ART of the limiting beltline material to determine the beltline pressure-temperature (P-T) limits to account for irradiation effect. The applicant stated that it will revise P-T limits for the four D/QCNPS units and submit them to the NRC for approval before the start of the extended period of operation using an approved fluence methodology. The applicant further stated that it will use ASME Code Cases N-640 and N-588 (Dresden Unit 3 only) for revising the P-T limits. The applicant will manage the P-T limits using approved fluence calculations when there are changes in the power of core design in conjunction with surveillance capsule results from the BWRVIP integrated surveillance program.

4-6 4.2.1.6 Reactor Vessel Circumferential Weld Examination Relief The analysis in BWRVIP-05, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, showed that the failure rate of the reactor vessel shell axial weld is orders of magnitude greater than 40-year end-of-life circumferential weld failure rate. This analysis has been used to justify relief from inspection of the circumferential welds. The circumferential weld examination relief analysis meets the requirements of 10 CFR 54.3(a) and is, therefore, a TLAA. The applicant presented the following data in LRA Table 4.2.6-1 for the circumferential welds of the Dresden vesselsmean values for copper and nickel contents for each vessel, mean 54-EFPY neutron fluence (E>1 MeV), delta RTNDT, and mean 54-EFPY RTNDT. In LRA Table 4.2.6-1, the applicant also presented the 64-EFPY data for circumferential welds of a Babcock and Wilcox (B&W) vessel from Table 2.6-5 of the final safety evaluation report (FSER) for BWRVIP-05. In a letter dated April 17, 2003, from Mr. P.R. Simpson, Exelon Nuclear, to the NRC, the applicant sent corrections to LRA Table 4.2.6-1. The corrected values of delta RTNDT and mean 54-EFPY RTNDT are 44 F and 54 F for Dresden Unit 2, and 58 F and 53 F for Dresden Unit 3. The comparison of the corrected Dresden data with the BWRVIP-05 FSER data indicates that the Dresden 54-EFPY mean RTNDT values are bounded by the 64-EFPY RTNDT value of 129 F for a B&W vessel. The applicant therefore indicated that the conditional failure probability of Dresden RPV circumferential welds is bounded by the corresponding NRC analysis results presented in the SER for BWRVIP-05. As indicated in LRA Section 4.2.6, this TLAA applies only to the two Dresden units because Dresden received relief from the reactor vessel circumferential weld examination for the remainder of the 40-year licensed operating period at the time the LRA was prepared. However, the applicant submitted a similar relief request for Quad Cities on May 16, 2003, which is currently being reviewed by the staff. The applicant was required to submit an update to LRA Section 4.2.6 to include the Quad Cities vessel circumferential weld examination relief analysis in accordance with 10 CFR 54.3(a) upon the staffs approval of the May 16, 2003, relief request. This was identified as Confirmatory Item 4.2.1.6. In response to Confirmatory Item 4.2.1.6, in a letter dated March 5, 2004, the applicant submitted a revision to the UFSAR supplement for the reactor vessel circumferential weld examination relief TLAA. The revised supplement refers to the documents related to RPV circumferential weld relief request extension for the license renewal term. The staff reviewed this supplement and found that it provides an adequate summary description regarding the evaluation of this TLAA. Therefore, Confirmatory Item 4.2.1.6 is closed.

4.2.1.7 Reactor Vessel Axial Weld Failure Probability As discussed in the preceding paragraph, Dresden has followed the BWRVIP-05 results and received relief from the circumferential weld inspections for the remaining 40-year licensed operating period. Quad Cities never submitted for this relief. Therefore, the analysis for the reactor vessel axial weld failure probability is a TLAA only for the two Dresden units. For the extended operating period, the applicant presented the limiting axial weld 54-EFPY properties for Dresden Units 2 and 3 in LRA Table 4.2.7-1. In a letter, dated April 17, 2003, from Mr. P.R.

Simpson, Exelon Nuclear, to the NRC, the applicant sent corrections to LRA Table 4.2.7-1.

The corrected values of delta RTNDT and mean 54-EFPY RTNDT are 44 F and 67 F for both Dresden Units 2 and 3. The corrected data are compared with the corresponding data for a B&W vessel from Table 2.6-5 in the NRC SER for BWRVIP-05 and the data for the Clinton plant from the supplement for the SER. The applicant stated that the Dresden limiting axial weld chemistry, chemistry factor, and the mean 54-EFPY RTNDT values are within the limits of

4-7 the values assumed in the analysis performed by the NRC staff in the March 7, 2000, BWRVIP-05 SER Supplement and the 64-EFPY limits and values presented in Table 2.6-5 of the SER.

The determination of the failure frequency of the limiting axial weld assumes that essentially 100 percent (i.e., 90 percent) of the vessel axial welds are inspected. However, because of various obstructions within the Dresden reactor vessels, the actual inspection can include less than 90 percent of axial welds. The applicant performed an analysis for the current 40-year operating period to assess the effect on the probability of fracture of the vessel axial welds because of the actual inspection performed. The analysis indicated that the conditional probabilities of failure because of a low-temperature over-pressurization event are very small, 3.89 x 10-8 and 5.07 x 10-8 on a per year basis for Dresden Units 2 and 3, respectively. The analysis results show that the calculated unit-specific axial weld conditional failure probabilities at 54 EFPYs are less than the failure probabilities calculated by the NRC staff for a B&W-fabricated vessel, presented in Table 2.6-5 of the BWRVIP-05 SER at 64 EFPYs, and the limiting Clinton values found in Table 3 of the SER Supplement. The applicant, therefore, indicated that the probability of failure of an axial weld at Dresden will provide adequate margin above the probability of failure of a circumferential weld, in support of relief from inspection of circumferential welds, for the extended operating period.

4.2.2 Staff Evaluation 4.2.2.1 Reactor Vessel Materials Upper-Shelf Energy Reduction Due to Neutron Embrittlement Section IV.A.1a of Appendix G of 10 CFR Part 50 requires, in part, that the RPV beltline materials have Charpy USE values in the transverse direction for base metal and along the weld for weld material of no less than 50 ft-lb (68J) throughout a facilitys license period, unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy USE will ensure margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

By letter dated April 30, 1993, the Boiling Water Reactor Owners Group (BWROG) submitted a topical report entitled 10 CFR Part 50 Appendix G Equivalent Margins Analysis for Low Upper Shelf Energy in BWR-2 Through BWR-6 Vessels, to demonstrate that BWR RPVs could meet margins of safety against fracture equivalent to those required by Appendix G to the ASME Code,Section XI, for Charpy USE values less than 50 ft-lb. In a letter dated December 8, 1993, the staff concluded that the topical report demonstrated that the evaluated materials (including BWR/3-6 plates, Non-Linde SAW welds, and electroslag welds (ESW))

have the margins of safety against fracture equivalent to Appendix G of ASME Code Section XI, in accordance with Appendix G of 10 CFR Part 50. In that report, the BWROG derived through statistical analysis the unirradiated USE values for materials that originally did not have documented unirradiated Charpy USE values. Using these statistically derived Charpy USE values, the BWROG predicted the end-of-license (40 years of operation) USE values in accordance with RG 1.99, Revision 2. According to this RG, the decrease in USE is dependent upon the amount of copper in the material and the neutron fluence at the 1/4T depth predicted for the material. The BWROG analysis determined through an equivalent

4-8 margins analysis (EMA) methodology that the minimum allowable Charpy USE value in the transverse direction for base metal and along the weld for weld material was 35 ft-lb.

General Electric updated the projected USE values for BWR RPV materials out to 54 EFPY in Electric Power Research Institute (EPRI) report TR-113596, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines, BWRVIP-74, September 1999. The staffs review and approval of EPRI TR-113596 is documented in a letter from Mr. C.I. Grimes to Mr. C. Terry dated October 18, 2001. The analysis in EPRI TR-113596 determined the reduction in the unirradiated Charpy USE resulting from neutron irradiation using the methodology in RG 1.99, Revision 2. Using this methodology, and a correction factor of 65 percent for conversion of the longitudinal properties to transverse properties, the lowest Charpy USE value for all BWR/3-6 plates was projected to be 45 ft-lb.

The correction factor for specimen orientation in plates is based on NRC Branch Technical Position MTEB 5-2. Using the RG methodology, the lowest Charpy USE value for BWR non-Linde 80 submerged arc welds was projected to be 43 ft-lb. According to EPRI TR-113596, the percent reductions in Charpy USE for the limiting BWR/3-6 beltline plates and BWR non-Linde 80 submerged arc welds are 23.5 percent and 39 percent, respectively.

Since the analysis in EPRI TR-113596 is a generic analysis, the applicant submitted plant-specific information in LRA Tables 4.2.1-1 through 4.2.1-4 for Dresden Units 2 and 3, and in LRA Tables 4.2.1-5 through 4.2.1-8 for Quad Cities Units 1 and 2, to demonstrate that the beltline materials of the D/QCNPS RPVs meet the criteria in the EPRI report at the end of the license renewal period. The tables include the information as specified in Tables B-4 and B-5 of EPRI TR-113596. In request for additional information (RAI) 4.2.1(a), the staff noted the applicants statement that it has performed one bounding 54-EFPY fluence calculation for Dresden and one for Quad Cities and then determined the corresponding 54-EFPY 1/4T fluence. Therefore, it was expected that the applicant used the same 54-EFPY 1/4T fluence for limiting beltline plate and weld material at both Dresden units. However, the data presented in Tables 4.2.1-1 through 4.2.1-4 indicate that the applicant used two different values for the limiting beltline materials for Dresdena fluence value of 3.9 x 1017 n/cm2 for the limiting plate and weld at Unit 2 and for the limiting plate at Unit 3, and a value of 2.9 x 1017 n/cm2 for the limiting weld at Unit 3. A similar apparent discrepancy was present in LRA Tables 4.2.1-5 through 4.2.1-8 for Quad Cities. In addition to this discrepancy, there appeared to be another discrepancy between the peak fluence data for Quad Cities in LRA Sections 4.2.1 and 4.2.2. Tables 4.2.1-5 through 4.2.1-7 of the LRA for Quad Cities list 2.9 x 1017 n/cm2 as the 54-EFPY 1/4T fluence, whereas LRA Table 4.2.2-2 for Quad Cities lists 3.9 x 1017 n/cm2 as the 54-EFPY 1/4T fluence. A similar discrepancy existed between LRA Sections 4.2.1 and 4.2.2 for the 1/4T fluence data for Dresden. In RAI 4.2.1(a), the staff requested the applicant to explain these apparent discrepancies in the peak fluence data and provide revised tables as appropriate.

In response to RAI 4.2.1(a) in a letter dated October 3, 2003, the applicant explained that the statement about the bounding 54-EFPY fluence calculation in LRA Section 4.2.1 meant that one neutron transport (flux) calculation was prepared that bounds both Dresden and Quad Cities. However, based upon the different operating bases for the four units with regard to the time period of operation at two different power levels, one before the extended power uprate (EPU) and one after EPU, a unit-specific fluence was calculated for each of the four units.

From these calculations, it can be seen that, using the bounding flux with the plant-specific pre-EPU and EPU periods of operation, the peak RPV fluence at 54 EFPY is the same (5.7 x

4-9 1017 n/cm2 at the RPV inside surface and 3.9 x 1017 n/cm2 at the 1/4T) for all four units, when rounding is applied. The staff has reviewed these values and found them acceptable.

The peak EPU fluence on the vessel is located at approximately 82 inches above the bottom of active fuel and is applied to the lower-intermediate shell and axial welds. Additionally, axial flux distribution factors are applied to different elevations (by shell) in the beltline region. For the lower shell, the peak fluence is adjusted by the axial flux distribution factor based on an elevation approximately 42 inches above the bottom of active fuel, which represents the lower to lower-intermediate girth weld. The axial flux distribution factor for this location is 0.71. The applicant stated that it applied this factor for calculating the peak pre-EPU fluence for the lower to lower-intermediate shell girth weld and all lower shell materials. In a followup question to RAI 4.2.1(a), the staff requested the applicant to describe how the pre-EPU axial flux profile compares with the EPU axial flux profile. The staff also requested that the applicant submit information about how the axial flux distribution factor was used in calculating the peak-EPU fluence for the lower to lower-intermediate shell girth weld and all lower shell materials. This was identified as Confirmatory Item 4.2.1(a).

In a letter dated April 9, 2004, the applicant referred to Figure 2 in a letter from Exelon to NRC, Additional Information Regarding Request for License Amendment for Pressure-Temperature Limits, dated July 31, 2003. This figure shows the pre-EPU and EPU axial flux distribution at the inside surface of the reactor pressure vessel. The pre-EPU and EPU axial flux distribution profiles are different, since the pre-EPU flux peaks at an elevation higher than the mid-plane, whereas the EPU flux peaks at the mid-plane. The applicant stated that for determining the peak 54-EFPY surface fluences at the lower shell plate material, lower shell welds and the lower to lower-intermediate shell girth weld, the axial flux distribution factor of 0.71 is applied for pre-EPU and 0.74 is applied for EPU conditions. The staff has independently verified the axial flux distribution factors using the data presented in the figure mentioned above and also verified the peak surface fluences for the lower shell and associated welds as calculated by the applicant. The staff finds the response acceptable because the applicant has used appropriate axial flux distribution factors for calculating the peak 54-EFPY surface fluence for the lower to lower-intermediate shell girth weld and all lower shell materials when determining the limiting materials. Therefore, Confirmatory Item 4.2.1(a) is closed.

In response to RAI 4.2.1(a), the applicant also provided the following explanation for the apparent discrepancies in the 54-EFPY fluence data for calculating the limiting beltline materials USE values presented in LRA Tables 4.2.1-1 through 4.2.1-8. In calculating the USE percent decrease for the limiting beltline material (plate or weld) of each unit, a combination of the applied fluence and the percentage of copper of each material is considered. Both the limiting plate and limiting weld materials for Dresden Unit 2 are in the lower-intermediate shell, thereby using the same fluence. The limiting plate material for Dresden Unit 3 also occurs in the lower-intermediate shell, thereby using the same fluence as that used for the Dresden 2 materials. However, the Dresden 3 limiting weld material with respect to the limiting USE percent decrease occurs in the lower to lower-intermediate girth weld because of the higher copper content, which offsets the higher fluence and lower copper content of the weld materials in the other shells. The Dresden Unit 3 lower to lower-intermediate girth weld sees a different (and lower) fluence than the lower-intermediate shell materials.

4-10 For Quad Cities Unit 1, the limiting plate and limiting weld materials with respect to 54-EFPY USE values occur in the lower shell, where the fluence is lower. In Quad Cities Unit 2, the limiting plate material is in the units lower shell, while the limiting weld material occurs in the lower-intermediate shell where the fluence is higher. The staff has independently verified the percentage of copper contents given in LRA Tables 4.2.1-1 to 4.2.1-8 for the limiting beltline USE materials with the corresponding data in the NRC Reactor Vessel Integrity Database (RVID). The staff accepts this explanation for the differences in the 54-EFPY fluence data for limiting beltline materials USE values because these values are determined by the different combinations of copper content and applied fluence for different beltline materials at D/QCNPS.

In further responding to RAI 4.2.1(a), the applicant provided the following explanation for the differences between the peak fluence data presented in LRA Sections 4.2.1 and 4.2.2 for Dresden and Quad Cities. The values presented in LRA Tables 4.2.2-1 and 4.2.2-2 represent the peak RPV fluence, both at the surface and at the 1/4T locations. As noted above, an axial flux distribution factor is applied to the lower shell, thereby reducing the fluence (both surface and 1/4T) for the associated materials. The values for delta RTNDT and ART provided in these tables represent the limiting materials based upon the fluence values presented. The staff finds the applicants explanation for the differences in fluence values acceptable because the fluences presented in Section 4.2.1 are for the limiting beltline material USE values, whereas those presented in Section 4.2.2 are for the limiting beltline material delta RTNDT and ART values.

The data for copper content in the limiting beltline plate and limiting beltline weld material presented in LRA Section 4.2.1 appear to be different from the data presented in Appendix F to the Dresden UFSAR. For example, LRA Table 4.2.1-2 lists 0.24 percent copper for the Dresden Unit 2 limiting beltline weld material, whereas Table 22 in Appendix F lists a maximum copper content of 0.21 percent for Dresden Unit 2. In RAI 4.2.1(b), the staff requested the applicant to resolve this apparent discrepancy. This was identified as Confirmatory Item 4.2.1.

In response to RAI 4.2.1(b), in a letter dated October 3, 2003, the applicant provided the following explanation:

For the beltline region, Table 21 (Shell Course 57Lower Shell) and Table 22 (Shell Course 58Lower-Intermediate Shell) of the Dresden FSAR gives values actual chemical analysis of these materials. Tables 21 and 22 contain the chemical analysis for electroslag welds contained in the original FSAR. Since the original publication of the FSAR, the accepted best estimate chemistry for Electroslag Weld (ESW) materials used in B&W vessels accepted by the NRC staff is 0.24% Cu and 0.37% Ni. These values are reported in BAW-2258, Evaluation of RTNDT, USE and Chemical Composition of Core Region Electroslag Welds for Dresden Units 2 and 3, Framatome Technologies, January 1996, and were previously accepted by the NRC in its review of pressure temperature (P-T) limit curve report GE-NE-B13-02057-04R1a. Exelon submitted reactor vessel chemistry values to the NRC in July 1998 in response to Generic Letter 92-01, Supplement 1. The information provided in that response is included in NRC database RVID.

The staff accepts the applicants response because, as mentioned above, the staff has verified the percentage of copper content given in LRA Tables 4.2.1-1 to 4.2.1-8 for the limiting beltline USE materials with the corresponding data in RVID. Therefore, Confirmatory Item 4.2.1 is closed.

4-11 In RAI 4.2.1(c), the staff requested the applicant to provide all fluence data for all welds and plates in the beltline and specify which one is bounding with respect to the RPV USE evaluation. In response to RAI 4.2.1(c), in a letter dated October 3, 2003, the applicant provided 54-EFPY surface fluences and 54 EFPY 1/4T fluences for all the beltline material but identified materials that are bounding with respect to the RPV material ART values at 54-EFPY. The applicant also needed to identify the USE for all beltline materials at 54-EFPYs and to identify the limiting materials for each unit. This was identified as Open Item 4.2.1(c).

Information was provided by the applicant in a letter dated April 9, 2004. The staff has reviewed this information and confirmed the limiting beltline materials for each unit. The staff also confirmed the USE for all four units and has reviewed the analysis for the material with the lowest Charpy USE, as described below.

The applicants April 9, 2004, letter indicated that all beltline materials, except for the ESWs in Quad Cities Unit 2, will have predicted Charpy USE greater than 35 ft-lb, the minimum allowable USE based on the generic BWR equivalent margins analysis documented in BWROG topical report entitled, 10 CFR Part 50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR-6 Vessels. Therefore, all beltline materials, except for the ESW in Quad Cities Unit 2, meet the margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

The applicant reevaluated the USE value for Quad Cities Unit 2 ESW using all electroslag weld material surveillance test results from Quad Cities Unit 2, and performed a plant-specific EMA for the Quad Cities Unit 2 ESW. General Electric report GE-NE-0000-0027-0575-01, Revision 0, The Upper Shelf Energy Evaluation for RPV Electroslag Welds at Quad Cities Unit 2, issued March 5, 2004, and included in the applicants April 9, 2004 letter, contains this analysis. Using the limiting surveillance capsule 18 data and the methodology in RG 1.99, Revision 2, the predicted Charpy USE for the ESWs welds is 34.2 ft-lb, which is below the minimum established in the generic BWROG topical report. The applicants plant-specific EMA was performed using methods and criteria contained in RG 1.161, Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy less than 50 Ft-Lb. and Appendix K of ASME Code,Section XI. Appendix K and RG 1.161 provide acceptance criteria and evaluation procedures for determining acceptability for operation of a reactor vessel when the vessel metal temperature is in the upper shelf range. The methodology is based on the principles of elastic-plastic fracture mechanics. Flaws will be postulated in the reactor vessel at locations of predicted low upper shelf Charpy impact energy, and the applied J-integral for these flaws will be calculated and compared with the J-integral fracture resistance of the material to determine acceptability. The applicants analysis showed that the applied J-integral of the postulated flaws and the J-integral material fracture resistance with a minimum USE of 32.4 ft-lb satisfies the criteria of Appendix K of the ASME Code,Section XI and RG 1.161.

The analysis methods in Appendix K of the ASME Code initially followed the methodology in RG 1.161. The analysis methods in Appendix K of the ASME Code,Section XI were changed in the 1995 Addenda to the 1995 Edition. The analysis method in the 1995 Addenda to the 1995 Edition of the ASME Code changed the method of calculating the contribution to the applied J-integral because of a radial thermal gradient. This change was incorporated into the ASME Code to more accurately represent the contribution to the applied J-integral due to a radial thermal gradient. The applicants analysis was performed using the earlier analysis method, i.e., the methods contained in RG 1.161. The staff confirmed the EMA using the analysis methods in both Appendix K to the ASME Code,Section XI, 1995 Addenda to the

4-12 1995 Edition, and the earlier analysis method in RG 1.161. This analysis included the effects of the extended power uprate condition. Since the limiting end of extended life USE for Quad Cities Unit 2 ESW exceeds the minimum value of 32.4 ft-lb demonstrated in the applicants plant-specific EMA, the staff concludes that all beltline materials, including the ESW in Quad Cities Unit 2 RPV meet the margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. Therefore Open Item 4.2.1(c) is closed.

4.2.2.2 Adjusted Reference Temperature for Reactor Vessel Materials Due to Neutron Embrittlement The applicant calculated the 54-EFPY fluences for the Dresden and Quad Cities reactor vessels using the methodology of NEDC-32983P. Because this methodology is approved by the NRC, the calculated 54-EFPY fluences are acceptable. The applicant provided the results for one bounding calculation and determined the peak surface fluence of 5.7 x 1017 n/cm2 and peak 1/4T fluence of 3.9 x 1017 n/cm2 for all four D/QCNPS vessels. Using the calculated peak 1/4T fluence, the applicant determined the 54-EFPY delta RTNDT and ART values for all the beltline materials according to RG 1.99, Revision 2. From all the 54-EFPY ART values, the applicant identified the limiting ART value and listed it in LRA Tables 4.2.2-1 and 4.2.2-2 as the 54-EFPY ART for both Dresden and Quad Cities. These limiting ART values are for axial welds. In RAI 4.2.2(a), the staff requested the applicant to explain how it determined that the weld in which it calculated the neutron fluence bounds all the other welds in Dresden/Quad Cities. In response to RAI 4.2.2(a), the applicant provided the same explanation for determining peak fluence that it had provided in response to RAI 4.2.1(a). The staff finds the applicants response acceptable because it has checked the applicants calculations for peak fluence and found them accurate.

In further responding to RAI 4.2.2(a), the applicant presented data for 54 EFPY 1/4T peak fluence and 54 EFPY ART for all beltline welds in Dresden and Quad Cities. The ESW in the lower-intermediate shell is a bounding weld material, one with the highest 54 EFPY ART, for each of the four Dresden and Quad Cities units. The staff has independently verified the identification of the bounding beltline material and its 54 EFPY ART at the 1/4 T location. The staff accepts the applicants response because the staff has verified it using the peak fluence data provided by the applicant and the percentage of copper and nickel contents, initial RTNDT, and margin data provided in the RVID. The applicant noted that the ART for the Dresden Unit 3 ESW material is 0.4 F less than that for the girth weld (lower to lower-intermediate weld) material. However, using ASME Code Case N-588, which allows a different application of KI for girth weld materials, the limiting material for Dresden Unit 3 is the ESW material, which is explained in more detail in the response to RAI 4.2.2 (c).

In RAI 4.2.2(b), the staff requested the applicant to submit the 54-EFPY delta RTNDT and ART values along with initial RTNDT for all the beltline materials of the four D/QCNPS reactor vessels. In response to RAI 4.2.2(b), in a letter dated October 3, 2003, the applicant provided the requested information in a table form for each of the four Dresden and Quad Cities units.

The staff finds the response acceptable because the information submitted by the applicant (percentage of copper and nickel, initial RTNDT, and margin term) is consistent with the data presented in the RVID and confirms the limiting materials.

The applicant stated that because of the refinement in the approved methodology used to calculate the 54-EFPY fluence, the material with the limiting ART is the axial weld, with the

4-13 exception of Dresden Unit 3, where the axial weld and girth weld ART values are identical.

The applicant invoked ASME Code Case N-588 for Dresden Unit 3 which causes the axial weld to become the limiting material. In RAI 4.2.2(c), the staff requested the applicant to identify the refinement mentioned here and explain how this makes the axial weld a material having the limiting ART. The staff also requested the applicant to explain how the use of ASME Code Case N-588 makes the axial weld the limiting material for Dresden Unit 3.

In response to RAI 4.2.2(c), in a letter dated October 3, 2003, the applicant stated that as can be seen in the ART tables submitted in response to RAI 4.2.2(b), the ART for the Dresden Unit 3 girth weld material is 104.16 F and the ART for the axial ESW material is 103.8F. The detailed explanation and calculated basis for the use of the ESW material as the limiting material is provided in GE-NE-0000-0002-9600-01a, Revision 0, and is explained in the GE Report. Because the calculated value of KIm is reduced for a girth weld because of the implementation of ASME Code Case N-588 (circumferentially oriented defect for a circumferential weld), the axial weld bounds the P-T curve beltline region requirements. The applicant also submitted the stress intensity calculations for both axial and girth welds at 54 EFPYs to demonstrate that by using ASME Code Case N-588 the axial weld has the most limiting temperature for the P-T curves in the beltline region. The results of the axial and girth weld calculations show that for a pressure of 1105 psig at 54 EFPYs, the allowable temperature (T) value for the axial weld (146.5 F) bounds the value for the girth weld (51 F).

The staff finds the response acceptable because the applicant has invoked ASME Code Case N-588, which has been approved by the staff as documented in the safety evaluations dated February 4, 2000 (ML003680441) for Quad Cities and August 25, 2000, (ML003745769) for Dresden.

4.2.2.3 Reflood Thermal Shock Analysis of the Reactor Vessel The applicant stated that the original D/QCNPS reflood thermal shock analysis has been superseded by an analysis for BWR-6 vessels that is applicable to the D/QCNPS BWR-3 reactor vessels. In RAI 4.2.3(a), the staff requested the applicant to explain why the BWR-6 analysis is applicable to the BWR-3 reactor vessel at D/QCNPS. In response to RAI 4.2.3(a),

in a letter dated October 3, 2003, the applicant stated that the BWR-6 evaluation determined the maximum stress intensity in the vessel wall as a function of vessel wall thickness and time after a design-basis LOCA. As shown in Figure G2214-1 of Appendix G to the ASME Code,Section XI, 1998 Edition through 2000 Addenda, the stress intensity is a function of vessel wall thickness. The original analysis used a recirculation line break, while the BWR-6 analysis was based on a main steam line break event, which is considered to bound the recirculation line break. In addition, the analysis used a vessel thickness similar to Dresden and Quad Cities vessels. Therefore, the BWR-6 analysis is applicable to the Dresden and Quad Cities reactor vessels. The staff finds the applicants explanation for applicability of an analysis for BWR-6 vessels to the D/QCNPS reactor vessels to be acceptable because the analysis for BWR-6 vessels bounds the recirculation line break event and it uses a vessel wall thickness similar to BWR-3 vessels.

The revised analysis assumes end-of-life material toughness, which in turn depends on end-of-life ART. The critical location for fracture mechanics analysis is at the 1/4T location. For the reflood thermal shock analysis of the reactor vessel, the peak stress intensity occurs at approximate 300 seconds after the LOCA. At that time, the temperature at 1/4T is approximately 204 C (400 F), which is much higher than the 54-EFPY ART 40 C (104 F)

4-14 for the limiting material of the D/QCNPS vessels. Therefore, the staff concurs with the applicant that the revised thermal shock analysis of the D/QCNPS vessels is valid for the period of extended operation.

4.2.2.4 Reflood Thermal Shock Analysis of the Reactor Vessel Core Shroud and Repair Hardware In the thermal shock analysis of D/QCNPS reactor vessel core shrouds, the applicant considered the location on the inside surface of the core shroud opposite to the midpoint of the fuel centerline as a location most susceptible to damage during an LPCI thermal shock transient because it receives the maximum irradiation. The 54-EFPY fluence at this location is 5.85 x 1020 n/cm2 (greater than 1 MeV). This fluence is calculated using the methodology of NEDC-32983P, which was approved by the NRC. In RAI 4.2.3(b), the staff requested the applicant to confirm whether the effect of extended power uprates, which is incorporated at D/QCNPS, is accounted for in the calculation of the 54-EFPY fluence.

In response to RAI 4.2.3(b), in a letter dated October 3, 2003, the applicant confirmed that the fluence used to determine the 54-EFPY shroud fluence was calculated using extended power uprate conditions. The staff finds the response acceptable because the calculations for the 54-EFPY shroud fluence take into account the effect of extended power uprates that are implemented at D/QCNPS.

The applicant calculated the maximum thermal shock stress and the corresponding thermal strain at the location on the inside surface of the shroud receiving the maximum irradiation.

The staff questioned the validity of this analysis. The reflood thermal shock would produce high tensile stresses on the outside surface of the core shroud, and these stresses would penetrate only to a small depth into the shroud wall. Thus, it appears that the outside surface of the core shroud could be the location most susceptible to damage during an LPCI thermal shock transient. In RAI 4.2.4-a, the staff requested the applicant to respond to the following three items:

(1) Provide an evaluation of strain at the outside surface of the core shroud, exposed to 54-EFPY fluence, during an LPCI thermal shock transient.

(2) What is the impact of strain rate associated with the LPCI thermal shock transient on the measured and calculated strains in the core shroud?

(3) The applicant compared the calculated strain range with the measured values of percent reduction in area for annealed Type 304 stainless steel irradiated to 1 x 1021 n/cm2 (E>1 MeV) and indicated that the analysis results represent a considerable margin of safety.

Provide the bases for concluding that the calculated strains at both the inside and outside surfaces of the shroud should be compared with the measured value of percent uniform strain for annealed Type 304 stainless steel irradiated to 1 x 1021 n/cm2 (greater than 1 MeV).

In response to RAI 4.2.4-a(1), in a letter dated October 3, 2003, the applicant submitted the calculation of the thermal shock strain for the LPCI transient described in the original analysis.

That analysis is based on a linear elastic thermal stress analysis and assumes that a low-pressure coolant of 49 C (120 F) is injected on a shroud at a temperature of 282 C

4-15 (540 F). The calculated thermal shock strain is 0.55 percent at the outside surface.

However, the effects of the thermal shock transient are very localized, and the majority of the material is at the higher temperature where the ductility is sufficient to accommodate the thermal shock strain and prevent brittle fracture. The staff finds the response acceptable because the localized thermal shock strain at the outside surface of the core shroud would be accommodated by the surrounding material at the higher temperature, thus preventing brittle fracture. This is further discussed in the next paragraph.

In response to RAI 4.2.4-a(2), in a letter dated October 3, 2003, the applicant stated that the thermal strains in the core shroud are calculated based on a linear elastic thermal stress analysis, which is a bounding calculation. The heat transfer coefficient is assumed to be infinite (making the calculation independent of strain rate), and therefore the outside surface of the shroud is considered to be at the fluid temperature 49 C (120 F). The applicant also stated that at the fluence levels experienced by the shroud, the material will continue to exhibit ductile behavior. As discussed above, the effect of the thermal shock transient is very localized, and the majority of the material is at the higher temperature where the ductility is sufficient to prevent brittle fracture. Even assuming the strain rate has a significant effect, the increased strain rate is still not sufficient to result in brittle fracture. The effect of strain rate during the LPCI thermal shock event can be accounted for by assuming that the material yield strength is increased (an effect also produced by increased fluence). At fluence levels up to 1 x1021 n/cm2 (which would represent the increased yield strength), Type 304 stainless steel exhibits sufficient ductility to preclude brittle fracture. The staff accepts the applicants assertion that Type 304 stainless has sufficient ductility at fluence levels up to 1x1021 n/cm2 because it is consistent with the data presented in a report by Mr. J.N. Kass, Effect of Neutron Irradiation at 288C (550F) on Reactor Component Materials for BWR-6, NEDO-20243, 74NED2, 1974. Kass reports that at a temperature of 288 C and neutron fluence of 1x1021 n/cm2 (>1MeV), Type 304 stainless steel experiences a 32 percent reduction in area. The staff accepts the applicant's response because the core shroud will have sufficient ductility during the LPCI transient during the extended period of operation to preclude brittle failure.

In response to RAI 4.2.4-a(3), in a letter dated October 3, 2003, the applicant stated that the strain associated with the reflood thermal shock event is very localized and is constrained by the surrounding bulk material. As such, it is similar to the triaxial stress condition present in the neck region (where the area reduction is taking place) during a tensile test. The percent reduction in area is a measure of this triaxial stress state and, as such, is the most appropriate property for evaluating the effect of thermal shock on the shroud. Therefore, a comparison with uniform elongation is not appropriate in this case. At lower values of temperature or neutron fluence, the percent reduction in area is generally higher. The staff accepts the applicant's reasoning that the strain associated with the reflood thermal shock event is very localized and, therefore, reduction in area is the most appropriate property for evaluating the effect of thermal shock on the shroud.

4.2.2.5 Reactor Vessel Thermal Limit AnalysesOperating Pressure-Temperature Limits The applicant plans to calculate vessel P-T limit curves for 60 years (54 EFPYs) using an approved fluence methodology for D/QCNPS and submit them to the NRC for approval before the start of the extended period of operation. The applicant will place the approved P-T limit curves in the D/QCNPS technical specifications. The applicant plans to use ASME Code Cases N-640 and N-588 (Dresden Unit 3 only) which have been approved by the staff as

4-16 documented in safety evaluations dated February 4, 2000 (ML003680441) for Quad Cities and August 25, 2000 (ML003745769) for Dresden. ASME Code Case N-640 allows the use of reference fracture toughness KIc, as found in Appendix A to ASME Code Section XI, in lieu of reference fracture toughness KIa, as found in Figure G-2210-1 in Appendix G to ASME Code,Section XI for the development of P-T limit curves. Reference fracture toughness KIc is based on the lower bound of static crack-initiation critical (reference) values of KI measured as a function of temperature. As found in Appendix G, KIa is based on the lower bound of static, dynamic, and crack arrest critical values of KI measured as a function of temperature. As mentioned above, the use of ASME Code Case N-588 for Dresden Unit 3 causes axial weld of the reactor vessel to become the limiting material. The applicant stated that it will manage the P-T limits using approved fluence calculations when there are changes in the power of core design in conjunction with surveillance capsule results from the BWRVIP integrated surveillance program. The staff finds the applicants plan to manage the P-T limits acceptable because the change in P-T curves will be implemented by the license amendment process (i.e., modifications of technical specifications) and will meet the requirements of 10 CFR 50.60 and Appendix G of 10 CFR Part 50.

4.2.2.6 Reactor Vessel Circumferential Weld Examination Relief Section 4.2.6 and Appendix A.3.1.6 to the LRA discuss inspection of the D/QCNPS RPV circumferential welds. These sections of the LRA indicate that the applicant will use an approved technical alternative in lieu of ultrasonic testing (UT) of RPV circumferential shell welds. The technical alternative is discussed in the staffs final SER of the BWRVIP-05 report, which is enclosed in a July 28, 1998, letter to Mr. C. Terry, the BWRVIP Chairman. In this letter, the staff concluded that since the failure frequency for circumferential welds in BWR plants is significantly below the criterion specified in RG 1.154, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors, and below the core damage frequency (CDF) of any BWR plant, the continued inspection would result in a negligible decrease in an already acceptably low value, and, therefore, elimination of the inservice inspection (ISI) for RPV circumferential welds is justified. The staffs letter indicated that BWR applicants may request relief from inservice inspection requirements of 10 CFR 50.55a(g) for volumetric examination of circumferential RPV welds by demonstrating that (1) at the expiration of the license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in the evaluation, and (2) the applicants have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the frequency specified in the report. The letter indicated that the requirements for inspection of circumferential RPV welds during an additional 20-year license renewal period would be reassessed, on a plant-specific basis, as part of any BWR LRA. Therefore, the applicant must request relief from inspection of circumferential welds during the license renewal period per 10 CFR 50.55a.

Section A.4.5 of the BWRVIP-74 report indicates that the staffs SER of the BWRVIP-05 report conservatively evaluated the BWR RPVs to 64 EFPYs, which is 10 EFPYs greater than what is realistically expected for the end of the license renewal period. The NRC staff used the mean RTNDT value for materials to evaluate failure probability of BWR circumferential welds at 32 and 64 EFPYs in the staff SER dated July 28, 1998. The mean RTNDT value is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT) and the mean value of the adjustment in reference temperature caused by irradiation (delta RTNDT); it does not include a

4-17 margin (m). The neutron fluence used in this evaluation was the neutron fluence at the clad-weld (inner) interface.

Since the staff analysis discussed in the BWRVIP-74 report is a generic analysis, the applicant submitted plant-specific information to demonstrate that the Dresden beltline materials meet the criteria specified in the report. To demonstrate that the Dresden vessels have not become embrittled beyond the basis for the technical alternative, the applicant, in LRA Table 4.2.6-1, supplied a comparison of 54-EFPY material data of the limiting Dresden circumferential welds with that of the 64-EFPY reference case in Appendix E to the staffs SER of the BWRVIP-05 report. The Dresden material data include the amounts of copper and nickel, chemistry factor, neutron fluence, delta RTNDT, initial RTNDT, and mean RTNDT of the limiting circumferential weld at the end of the renewal period. The staff has verified the data for the amounts of copper and nickel contents and the initial RTNDT values for Dresden Unit 2 and 3 beltline materials by comparing them with the corresponding data in the RVID. The 54-EFPY mean RTNDT values for Dresden Units 2 and 3 are 54 F and 53 F, respectively. The staff has checked the applicants calculations for the 54-EFPY mean RTNDT values for the Dresden circumferential welds using the data presented in LRA Table 4.2.6-1 and found them to be accurate. These 54-EFPY mean RTNDT values for Dresden Units 2 and 3 are bounded by the 64-EFPY mean RTNDT value of 129.4F used by the NRC for determining the conditional failure probability of a circumferential girth weld. The 64-EFPY mean RTNDT value from the staff SER dated July 28, 1998, is for a B&W weld because B&W welded the girth welds in the Dresden vessels. Since the Dresden 54-EFPY mean RTNDT values are less than the 64-EFPY value from the staff SER dated July 28, 1998, the staff concludes that the Dresden RPV conditional failure probabilities are bounded by the NRC analysis.

The applicant stated that the procedures and training used to limit cold over-pressure events will be the same as those approved by the NRC when Dresden requested to use the BWRVIP-05 technical alternative for the current term, but it did not explicitly cite a document that supports this statement. In RAI 4.2.6, the staff requested the applicant to provide specific reference(s) in the LRA and the UFSAR Supplement that include the applicants request to use the BWRVIP-05 technical alternative for the current license term and the NRC approval of that request. In response to RAI 4.2.6, in a letter dated October 3, 2003, the applicant stated that the procedure and training requirements identified in the Dresden request to use the BWRVIP-05, technical alternative were identified in Dresden Letter JMHLTR 99-0078 from J. M. Heffley (Commonwealth Edison (ComEd)) to NRC, Relief Request for Alternative Weld Examination of Circumferential Reactor Pressure Vessel Shell Welds, dated July 26, 1999, attached to Dresden ISI Relief Request No. CR-18. The NRC approval of this relief request and associated procedure and training requirements was provided in the document, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Alternative to Inspection of Reactor Pressure Vessel Circumferential Welds, Dresden Power Station, Units 2 and 3, attached to the NRC letter from Mr. A.J. Mendiola to Mr/ O.D. Kingsley (ComEd), Dresden, Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential Weld Examinations, (Task Action Commitment Nos. MA6228 and MA6229), dated February 25, 2000. The applicant further stated that LRA Section 4.2.6 and associated UFSAR Supplement Section A.3.1.6, Reactor Vessel Circumferential Weld Examination Relief, should have referenced the request letter identified above. By letter dated March 5, 2004, the applicant provided revisions to the UFSAR Supplements A.3.1.6 for Dresden and Quad Cities which clarified that the application for relief was extended to the period of extended operation. The staff finds the response acceptable because the applicant identified the requested references.

4-18 With respect to Quad Cities, the applicant submitted its request for relief from the circumferential welds examination requirements for the remainder of the 40-year licensed operating period by letter dated May 16, 2003. This relief was approved by the NRC staff by SER dated April 29, 2004. However, the applicant did not provide an evaluation of the basis for extending this relief through the end of the period of extended operation, as was done for Dresden. The applicant submitted an extension of the relief request for all four units through the period of extended operation. The staff will review the basis for extending this relief through the end of the period of extended operation for all four units as the staff completes its review of the applicants February 23, 2004 submittal.

The staff finds that the applicants evaluation for this TLAA is acceptable because the Dresden 54-EFPY conditional failure probabilities for the RPV circumferential welds are bounded by the NRC analysis in the staff SER dated July 28, 1998, and Dresden will be using procedures and training to limit cold over-pressure events during the period of extended operation. This analysis satisfies the evaluation requirements of the staff SER dated July 28, 1998. The applicant has provided this request for relief for the extended period of operation by letter dated February 23, 2004 for Dresden and Quad Cities and will be reviewed by the staff in accordance with 10 CFR 50.55a. The staff finds the UFSAR Supplement (LRA Appendix A.3.1.6) for Dresden and Quad Cities acceptable because it includes the necessary information regarding the evaluation of this TLAA.

4.2.2.7 Reactor Vessel Axial Weld Failure Probability In its July 28, 1998, letter to Mr. C. Terry, the BWRVIP Chairman, the staff identified a concern about the failure frequency of axially oriented welds in BWR RPVs. In response to this concern, the BWRVIP supplied evaluations of axial weld failure frequency in letters dated December 15, 1998, and November 12, 1999. The staffs SER on these analyses is enclosed in a March 7, 2000, letter to Mr. C. Terry. The SER indicates that the RPV failure frequency because of failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is below 5 x 10-6 per reactor year, given the assumptions about flaw density, distribution, and location described in this SER. Because the results apply only for the initial 40-year license period of BWR plants, applicants for license renewal must submit plant-specific information applicable to 60 years of operation.

The BWRVIP identified the Clinton and Pilgrim reactor vessels as the reactor vessels with the highest mean RTNDT in the BWR fleet. The mean RTNDT values were determined using the neutron fluence at the clad/weld interface and did not include a margin term. The staff confirmed this conclusion about the highest mean RTNDT in the SER enclosed in the March 7, 2000, letter by comparing the information in the BWRVIP analysis and the information in the RVID for all BWR RPV axial welds. The results of the staff and BWRVIP calculations are presented in Table 4.2-1. The staff calculations used the basic input information for Pilgrim, with three different assumptions for the initial RTNDT. The calculations of the actual Pilgrim condition used the docketed initial RTNDT of -44 C (-47 F) and a mean RTNDT of 20 C (68 F). A second calculation, listed as Mod 1 in Table 4.2-1, uses an initial RTNDT of -18 C (0 F) and a mean RTNDT of 47 C (116 F) consistent with the BWRVIP calculations. A third calculation, with an initial RTNDT of -19 C (-2 F) and a mean RTNDT of 46 C (114 F), was chosen to identify the mean value of RTNDT required to provide a result which closely matches the RPV failure frequency of 5 x 10-6 per reactor year. The vessel failure frequency is the product of conditional failure probability, or P(F/E), and the low-temperature over

4-19 pressurization (LTOP) event frequency. The LTOP frequency is the frequency of the transient occurring, determined as 10-3 per reactor year in the evaluation. Since Pilgrim was not identified within the RVID as one of the limiting plants, the SER states that the axial welds for the Clinton reactor vessel are the limiting welds for the BWR fleet, and vessel failure probability determined for Clinton should bound those for the BWR fleet.

Table 4.2-1. Comparison of Results from Staff and BWRVIP Calculations Plant Initial RTNDT

(F)*

Mean RTNDT

(F)

Vessel Failure Freq.

Staff BWRVIP Clinton

-30 91 2.73E-6 1.52E-6 Pilgrim

-48 68 2.24E-7 Mod 1 **

0 116 5.51E-6 1.55E-6 Mod 2 ***

-2 114 5.02E-6

C = 0.56 x (F - 32)

- A variant of Pilgrim input data, with initial RTNDT = 0F

- - A variant of Pilgrim input data, with initial RTNDT = -2F Since the BWRVIP analysis was generic, the applicant submitted plant-specific information in LRA Section 4.2.7 to demonstrate that the Dresden beltline materials meet the criteria specified in the March 7, 2000, SER. The Dresden vessels were fabricated by B&W, and its axial welds are ESW. To demonstrate that the Dresden vessels have not become embrittled beyond the basis for the staff and BWRVIP analyses, the applicant submitted the following information in LRA Table 4.2.7-1:

a comparison of the amounts of copper and nickel, chemistry factor, neutron fluence, delta RTNDT, initial RTNDT, and mean RTNDT of the limiting axial welds at the end of the renewal period to the reference cases in the BWRVIP and staff analyses estimates of the conditional failure probability of the Dresden RPVs at the end of the license renewal term based on the comparison of the mean RTNDT for the limiting axial welds and the reference cases Table 4.2.7-1 of the LRA includes data for two reference cases, 64-EFPY data for a limiting B&W vessel (from Table 2.6-5 in the March 7, 2000, SER) and Clinton data from Table 4.2-1.

The data in LRA Table 4.2.7-1 show that the mean RTNDT values for Dresden Units 2 and 3 are equal to 19 C (67 F), and these values are smaller than the corresponding values for Clinton and the limiting B&W vessel. The staff verified the data for chemical composition and initial RTNDT for all the axial welds in the two Dresden vessels by comparing them with the corresponding data in RVID maintained by the NRC and confirmed the identification of the most limiting axial weld as presented in LRA Table 4.2.7-1. The staff checked the applicants calculations for mean RTNDT values following Equation 2 of RG 1.99, Revision 2, and found them to be accurate.

4-20 The vessel failure frequency calculated by the SER for BWRVIP-05 and its supplement assumes that essentially 100 percent of the axial welds can be inspected. According to 10 CFR 50.55.a(g)(6)(ii)(A)(2), essentially 100 percent as used in Table IWB-2500-1 means more than 90 percent of the examination volume of each weld, where the reduction in coverage results from interference by another component, or part geometry. However, the actual inspection at Dresden can include less than 90 percent of the axial welds. Therefore, the applicant performed an analysis to calculate the conditional probability of vessel failure, taking into account the actual limited inspection of axial welds at Dresden. The results show that the conditional probability of vessel failure at 54 EFPYs because of an LTOP event is very small, 3.89 x 10-8 and 5.07 x 10-8 for Dresden Units 2 and 3, respectively. These values of the conditional probability of failure of Dresden axial welds are smaller than the corresponding values calculated by the NRC staff in the SER at 64 EFPYs and the limiting Clinton values found in Table 4.2-1 of this SER. In RAI 4.2-7, the staff requested the applicant to confirm whether the analysis was performed as part of relief from 100 percent axial and/or elimination of circumferential inspection. The staff also requested the applicant to discuss the impact of 54 EFPYs of operation on the probability of vessel failure. In response to RAI 4.2-7, in a letter dated October 3, 2003, the applicant stated that this analysis was performed to demonstrate that the reliability of the Dresden RPVs remained extremely high considering actual inspection coverage. The actual inspection coverage could not meet the essentially 100 percent coverage because of inspection limitations caused by obstructions with internal components and attachments. The analysis did not include the circumferential welds since they had been previously eliminated from the inspection plan in accordance with BWRVIP-05. The applicant also stated that the failure probabilities quoted in the question were determined using the predicted fluence at the end of 60 years of operation (54 EFPYs). The staff finds the response acceptable because the failure probabilities listed in LRA Table 4.2.7-1 were calculated taking into account the actual inspection coverage and the predicted 54-EFPY fluence.

This axial weld failure probability analysis is required to be performed as a license renewal action item in accordance with the staff FSER of EPRI report TR-113596 (BWRVIP-74) and compliance with the license renewal rule (10 CFR Part 54) enclosed in an October 18, 2001, letter from Mr. C.I. Grimes to Mr. C. Terry. This action item, as stated in the staffs March 7, 2000, letter to Mr. C. Terry, requires the license renewal applicant to monitor axial beltline weld embrittlement. One acceptable method is to determine that the mean RTNDT of the limiting axial beltline weld at the end of the extended period of operation is less than the values specified in Table 1 of this FSER. Therefore this evaluation applies to Dresden Units 2 and 3, as well as to Quad Cities Units 1 and 2. In addition, Dresden and Quad Cities have the same mean RTNDT, because the initial RTNDT, chemical composition, and 54-EFPY surface fluence are the same for the limiting beltline axial welds at Quad Cities and Dresden. Therefore, for Quad Cities and Dresden plants, the mean RTNDT for the limiting beltline axial welds at 54-EFPYs is equal to 19 C (67 F). A comparison of the mean RTNDT value of 33 C (91 F) for the Clinton axial weld from Table 4.2-1 of this SER with the Dresden and Quad Cities value of 19 C (67 F) shows that the NRC analysis of the Clinton axial welds bounds the Dresden and Quad Cities welds. The applicant should confirm that Quad Cities Units 1 and 2 have a mean value of 19 C (67 F) and address this TLAA of the axial welds for Quad Cities in the UFSAR Supplement. This was identified as Confirmatory Item 4.2.2. In response to Confirmatory Item 4.2.2, in a letter dated March 25, 2004, the applicant compared the limiting axial weld 54-EFPY properties for Quad Cities 1 and 2 against the corresponding limiting values calculated by the NRC in the SER for BWRVIP-05 at 64 EFPY and the limiting Clinton values taken from Table 2.6-5 in the March 7, 2000, supplement to the SER. The applicant confirmed that the limiting

4-21 axial welds at Quad Cities Units 1 and 2 have a mean 54 EFPY RTNDT of 19ºC (67ºF), which is less than the value of 33ºC (91ºF) for Clinton. The comparison also shows that the conditional vessel failure probabilities for Quad Cities Units 1 and 2 are equal to 2.08 x 10-7 and 5.27 x 10-7, respectively. These failure probabilities are less than the corresponding value for Clinton listed in Table 4.2-1 of this SER. The staff finds the applicant's evaluation for this TLAA acceptable because the conditional probability of failure of Quad Cities Unit 1 and 2 limiting axial welds at 54 EFPY is smaller than the corresponding values calculated by the NRC staff in the SER for BWRVIP-05 at 64 EFPY and the limiting Clinton values found in the March 7, 2000, supplement to the SER.

In a letter dated March 5, 2004, the applicant submitted a revision to the UFSAR supplement for the reactor vessel axial weld failure probability. The staff reviewed this supplement and found that it provides an adequate summary description regarding the evaluation of this TLAA.

Therefore, Confirmatory Item 4.2.2 is closed.

The staff finds that the applicants evaluation for this TLAA is acceptable because the conditional probability of failure of Dresden and Quad Cities axial welds at 54 EFPYs is smaller than the corresponding values calculated by the NRC staff in the SER at 64 EFPYs and the limiting Clinton values found in the March 7, 2000, SER. The staff finds the UFSAR Supplement (LRA Appendix A.3.1.7) to be acceptable because it includes the necessary information regarding the evaluation of this TLAA.

4.2.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) that the analyses remain valid for the period of extended operation, which includes the extended power uprate conditions, for the reflood thermal shock analysis of the reactor vessel and the reflood thermal shock analysis of the reactor vessel core shroud and repair hardware TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the reflood thermal shock analysis of the reactor vessel and the reflood thermal shock analysis of the reactor vessel core shroud and repair hardware TLAA evaluations for the period of extended operation, as required by 10 CFR 54.21(d).

On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(ii), that the analyses have been projected to the end of the period of extended operation, which includes the extended power uprate conditions, for the reactor vessel materials USE reduction due to neutron embrittlement, adjusted reference temperature for reactor vessel materials due to neutron embrittlement, reactor vessel thermal limit analyses, reactor vessel circumferential weld examination relief, and the reactor vessel axial weld failure probability TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the reactor vessel material USE reduction because of neutron embrittlement, adjusted reference temperature for reactor vessel materials because of neutron embrittlement, reactor vessel thermal limit analyses, reactor vessel circumferential weld examination relief, and the reactor vessel axial weld failure probability TLAA evaluations for the period of extended operation, as required by 10 CFR 54.21(d).

4-22 Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.3 Metal Fatigue A metal component subjected to cyclic loading at loads less than the static design load may fail because of fatigue. Metal fatigue of components may have been evaluated based on an assumed number of transients or cycles for the current operating term. The validity of such metal fatigue analysis is reviewed for the period of extended operation. NUREG-1801 identifies fatigue aging-related effects that require evaluation as possible TLAAs pursuant to 10 CFR 54.21(c). Each of these is summarized in NUREG-1800 and presented in Section 4 of the LRA.

4.3.1 Summary of Technical Information in the Application The applicant discussed the design requirements for components of the reactor coolant system (RCS) at Dresden and Quad Cities. The reactor vessel, reactor vessel internals, and the reactor coolant pressure boundary (RCPB) piping and components were designed and fabricated in accordance with the requirements for Class 1 components stated in ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition and 1965 Summer Addenda. Other safety-related piping and fittings were designed and fabricated in accordance with the requirements of United States of America Standard (USAS) B31.1, Power Piping Code, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C. The fatigue analyses of both the reactor coolant loop and attached piping were performed in accordance with the requirements for ASME Code,Section III, Class 1 components.

4.3.1.1 Reactor Vessel Fatigue Analyses In Section 4.3.1 of the LRA, Reactor Vessel Fatigue Analyses, the applicant stated that the original pressure vessel stress report included ASME Code Section III fatigue analyses of the reactor vessel components based on a set of design-basis duty cycles, which are listed in Table 3.9-1 of the Dresden and Quad Cities UFSARs. The analyzed components consisted of the vessel support skirt, shell, upper and lower heads, closure flanges, nozzles and penetrations, and closure studs. The original 40-year analyses demonstrated that the cumulative usage factors (CUFs) for the critical components are below the ASME Code,Section III limiting value of 1.0. A reanalysis was performed for reactor vessel CUFs as part of EPU implementation at all four Dresden and Quad Cities units. A subset of the bounding reactor vessel components was evaluated as a part of this reanalysis. The current bounding-case analysis (worst CUFs for all four reactor vessels) lists the following values for the 40-year CUFs for the limiting components:

shroud support 0.820 support skirt 0.862 feedwater nozzle (safe end) 0.748 closure studs 0.750

4-23 The original code analysis of the reactor vessel included fatigue analyses of the feedwater nozzles and the control rod drive hydraulic system return line nozzles. These nozzles were found to be susceptible to cracking caused by a number of factors, including rapid thermal cycling. The control rod drive hydraulic system return line nozzles were therefore capped and removed from service. A reanalysis was also performed on the feedwater nozzles and modifications implemented to reduce or eliminate the effects of the high thermal cycling.

The applicant stated that the fatigue of the reactor vessel, including the support skirt, shell, upper and lower heads, closure assembly, nozzles and penetrations, and nozzle safe ends, will be managed by the Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program (Metal Fatigue AMP). The program is discussed in Section B.1.34 of the LRA. All governing reactor vessel fatigue analyses have been reviewed to establish a bounding set of reactor locations for inclusion in the Metal Fatigue AMP. Eight locations where the 40-year CUFs are expected to exceed the threshold value of 0.4 will be included in the program. These locations are listed in Table 4.3.1-1 of the LRA and include the four locations listed in the bullets above. The other locations are on the recirculation outlet nozzle, the recirculation inlet nozzle, the core spray nozzle, and the vessel shell. These components were selected because they are listed in NUREG/CR-6260. The feedwater nozzle safe end was also selected because it is one of the components listed in NUREG/CR-6260.

The applicant stated that Dresden and Quad Cities have installed programs to track thermal and pressure cycles and to assess their effect on vessel fatigue. The requirements from these procedures will be incorporated into the Metal Fatigue AMP. All necessary plant transient events will be tracked to ensure that the CUF remains less than 1.0 for all monitored components. In the event that the CUF for any component is projected to exceed 1.0 during the period of extended operation, appropriate corrective action will be taken in accordance with the Exelon Corrective Action Program. This program is discussed in Section B.2.1 of the LRA.

4.3.1.2 Fatigue Analysis of the Reactor Internals In Section 4.3.2 of the LRA, Fatigue Analysis of Reactor Vessel Internals, the applicant stated that a review of the CLBs for Dresden and Quad Cities identified only two fatigue analyses of reactor vessel internals.

In Section 4.3.2.1 of the LRA, Low Cycle Thermal Fatigue Analysis of the Core Shroud and Repair Hardware, applicable to Quad Cities only, the applicant identified in the CLB an evaluation for low-cycle mechanical fatigue of the core shroud and the rod stabilizers in the core shroud repair hardware because of a cold feedwater transient. The CUF for the core shroud was found to be negligible, while the CUF for the rod stabilizers for a 40-year plant life was calculated as less than 0.11. The applicant indicated that the design of the core shroud support hardware for fatigue effects is valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

In Section 4.3.2.2 of the LRA, High-Cycle Flow-Induced Vibration Fatigue Analysis of Jet Pump Riser Braces, applicable to Dresden Unit 2 only, the applicant identified core flow or single recirculation loop operation as a cause of significant vibration levels of reactor vessel internal components. To address this concern, the Dresden Unit 2 reactor vessel internals were instrumented and tested for vibration levels during startup of the plant. Limiting criteria were established such that vibration stress levels were assured to remain below material

4-24 endurance limits over the life of the plant. Various operating conditions were evaluated, including those associated with increased core flow and transient unbalanced flow conditions.

The limiting components were found to be the jet pump riser braces. Although reactor internals are not code pressure boundary components, the evaluation of the Dresden Unit 2 jet pump riser braces used methods and fatigue curves similar to those of ASME Code,Section III, Class 1 fatigue analyses.

The applicant stated that the EPU analyses found that Dresden and Quad Cities reactor internal components, with the exception of the Dresden Unit 2 jet pump riser braces, can operate at EPU conditions for the period of extended operation without exceeding the original design vibration criteria or developing increased vibration levels because of recirculation pump vane passing frequencies. The EPU project evaluated possible effects of the power uprate and found that, with some possible exceptions, including the Dresden Unit 2 riser braces, the stress ranges of the reactor internals would remain within the original endurance limit.

The EPU evaluation of flow-induced vibration of reactor internals found that the Dresden Unit 2 jet pump riser braces might be damaged by the recirculation pump vane passing frequency vibration if operation is permitted in the maximum extended load line limit analysis region.

Operation in this region might produce fatigue cracks and failure in the riser braces at Dresden Unit 2 only. The Dresden Unit 2 jet pump riser braces are susceptible to resonance effects, whereas the jet pump riser braces at Dresden Unit 3 and Quad Cities have shown no such effects and therefore present no concern.

The applicant stated in the LRA that the Dresden Unit 2 riser braces will be repaired or replaced before the start of the period of extended operation, and, pursuant to 10 CFR 54.21(c)(1)(ii), will be qualified for the period of extended operation.

4.3.1.3 ASME Code,Section III, Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis The applicant stated in Section 4.3.3.1 of the LRA, ASME Section III Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis, that the Dresden Unit 3 recirculation system piping is the only RCPB piping that was analyzed for fatigue per ASME Code,Section III, Class 1 rules at either Dresden or Quad Cities. The Dresden Unit 3 recirculation piping was replaced under the Generic Letter (GL) 88-01 intergranular stress-corrosion cracking mitigation program. The analysis included portions of the connected shutdown cooling system, the low-pressure coolant injection system, the isolation condenser system, and the reactor water cleanup system. All other Class 1 piping at both plants was initially designed to USAS B31.1, 1967 Edition.

The applicant stated that, in accordance with 10 CFR 54.21(c)(1)(iii), the Dresden Unit 3 RCPB piping will be managed by the Metal Fatigue AMP. This program is described in Section B.1.34 of the LRA. The program will monitor CUFs through the cycle-based fatigue monitoring option. All Dresden Unit 3 RCPB piping fatigue analyses have been evaluated to establish a bounding set of piping locations for inclusion in the Metal Fatigue AMP. All locations where the 40-year CUFs are expected to exceed a threshold value of 0.4 will be included in the AMP.

The ASME Code,Section III, Class 1 analyses for the Dresden Unit 3 recirculation line and attached large-bore piping replacement inside the drywell show calculated CUFs at seven locations that exceed the 0.4 threshold. These locations are listed in Table 4.3.3.3-1 of the

4-25 LRA. All applicable plant transient events will be tracked to ensure that the CUF remains less than the ASME Code,Section III, Class 1 fatigue limit of 1.0 at the monitored locations. In the event the CUF at a location is predicted to exceed 1.0 before 60 years of operation, the applicant stated that the necessary corrective action will be taken in accordance with the Exelon Corrective Action Program, described in Section B.2.1 of the LRA. The required implementing actions will be completed before the period of extended operation. The requirements of these procedures will also be incorporated into the Metal Fatigue AMP.

4.3.1.4 Reactor Coolant Pressure Boundary Piping and Components Designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C In Section 4.3.3.2 of the LRA, Reactor Coolant Pressure Boundary Piping and Components Designed to USAS B31.1, ASME Section III Class 2 and 3, or ASME Section VIII Class B and C, the applicant stated that the RCPB and non-RCPB piping for the Dresden and Quad Cities units, including the portions of the main steam and safety relieve valve (SRV) discharge lines inside the drywell, was designed to USAS B31.1 except for the replaced RCPB piping in Dresden Unit 3, described in Section 4.3.3.1 of the LRA. None of these codes requires explicit fatigue analysis. However, the RCPB and non-RCPB piping within the scope of license renewal that is designed to USAS B31.1 or ASME Code,Section III, Classes 2 and 3 requires the application of a stress reduction factor to the allowable thermal bending stress range if the number of full-range cycles exceeds 7000. The applicant also stated that, with the exception of containment vent and process bellows, no components within the scope of license renewal designed to ASME Code,Section III or Section VIII require design, or were designed, for thermal cycling. This applies to the reactor recirculation pumps, which were designed per ASME Code,Section III, Class C (1965), and the Quad Cities residual heat removal (RHR) system heat exchangers, which were designed to ASME Code,Section III, Class C (1996) requirements on the shell side and Section VIII on the tube side. The RHR system includes no TLAAs other than the piping design for USAS B31.1 stress range reduction factors.

The applicant indicated that the assumed thermal cycle count for the analyses can be approximated by the thermal-cycles used in the reactor vessel fatigue analysis. These thermal cycles are listed in UFSAR Table 3.9-1. The total count of all these listed thermal cycles is less than 2200 over the 40-year plant life. For the 60-year extended operating period, the number of assumed operating cycles would be increased to 3300, considerably less than the 7000-cycle threshold in USAS B31.1. In accordance with 10 CFR 54.21(c)(1)(i), the applicant indicated that the existing piping analyses within the scope of licence renewal containing assumed thermal-cycle counts are valid for the period of extended operation.

4.3.1.5 Fatigue Analysis of the Isolation Condenser In Section 4.3.3.3 of the LRA, Fatigue Analysis of the Isolation Condenser, the applicant stated that the Dresden isolation condensers (which provide core cooling when the reactor vessel becomes isolated from the main turbine and the main condenser) were initially designed for 280 thermal isolation operations. The ASME Code,Section III, Class 1 fatigue analysis of the critical components of the condensers determined that the 40-year CUF is below the ASME Code Section III limiting value of 1.0.

4-26 Based on the number of reactor scrams experienced since the start of operation and the number of recorded isolation operations, the applicant determined that the number of expected isolation condenser operations would be 181 through the 60-year extended period of operations. This projected cycle count is below the 280 isolation condenser operating design limit. On this basis the applicant indicates that, according to 10 CFR 54.21(c)(1)(i), the design fatigue analysis remains valid through the period of extended operation.

4.3.1.6 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)

In Section 4.3.4 of the LRA, the applicant described the actions taken to address the issue of environmentally assisted fatigue. Generic Safety Issue (GSI) 190 addresses the effects of reactor coolant environment on fatigue life of components and piping. Although GSI 190 is resolved, Section 4.3.1.2 of NUREG-1800 states that for license renewal, the applicants consideration of the effects of coolant environment on component fatigue life is an area of review.

The applicant stated that plant-specific calculations will be performed for Dresden and Quad Cities for the following fatigue sensitive component locations identified in NUREG/CR-6260 for older vintage BWRs:

reactor vessel (lower head to shell transition) feedwater nozzle recirculation system (RHR return line tee) core spray system (nozzle and safe end) residual heat removal line (tapered transition) limiting Class 1 location in a feedwater line This list does not specifically include the feedwater line reactor core isolation cooling (RCIC) tee location identified in NUREG/CR-626, because Dresden does not have an RCIC system and the RCIC tee location in Quad Cities is located in the outside containment in the Class 2 portion of the feedwater line. However, the applicant has committed to evaluate for each plant the limiting Class 1 feedwater piping location as stated above.

The applicant stated that for each location listed above, detailed environmental fatigue calculations will be performed using the appropriate Fen relationships from NUREG/CR-6583 Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, for carbon and alloy steels and those from NUREG/CR-5704 Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, for stainless steel, as appropriate for the material. These calculations will be completed before the period of extended operation, and appropriate corrective action will be taken if the resulting end-of-life CUF values exceed 1.0. The applicant also stated that it reserves the right to modify this position in the future, based on the results of ongoing industry activities on this topic, subject to NRC approval before changes in this position.

4-27 4.3.2 Staff Evaluation 4.3.2.1 Reactor Vessel Fatigue Analysis The applicant has identified CLB fatigue analyses associated with the reactor vessels as TLAAs, in conformance with the provisions of 10 CFR 54.21(c)(1) and the components listed in the appropriate tables in NUREG-1801. The applicant listed the 40-year bounding CUFs associated with these TLAAs and indicated that the CUFs for several locations exceeded a threshold value 0.4. These locations therefore have the potential to exceed the limiting value of 1.0 during the period of extended operation. The staff reviewed these locations and determined that they conform to the locations or components listed in NUREG-1801. The applicant has therefore committed to monitor, as part of the Metal Fatigue AMP, eight locations where the CUFs are expected to exceed the threshold value during the 40-year plant life. This is Commitment #34 in Appendix A of this SER. The applicant stated that as part of the Metal Fatigue AMP, all necessary plant transient events will be tracked and the CUFs calculated, to verify that the CUF remains less than 1.0 for all monitored components and locations. In the event that the CUF for any component is projected to exceed 1.0 before 60 years of operation, the applicant commits to take appropriate corrective action in accordance with the Exelon Corrective Action Program, described in Section B.2.1 of the LRA. This is Commitment #39 in Appendix A of this SER. The staff finds this acceptable because, pursuant to 10 CFR 54.21(c)(1)(iii), the applicant has provided assurance that an adequate margin of safety for the reactor vessel will be maintained for the period of extended operation, as reflected in its commitment to meet the ASME Code,Section III, Class 1 fatigue analysis criterion associated with the Metal Fatigue AMP, or to implement corrective actions associated with the Exelon Corrective Action Program. The applicants supplements for the Dresden and Quad Cities UFSARs regarding the reactor vessel fatigue analyses are provided in Section A.3.2.1 of the respective LRAs. The staff has reviewed these supplements and finds them acceptable because they provide a reasonable summary of the information presented in Section 4.3.1 of the LRA.

4.3.2.2 Fatigue Analysis of the Reactor Internals The staff has reviewed the sections of the LRA pertaining to the fatigue of reactor internals.

Based on a 40-year highest CUF of 0.11 for Quad Cities core shrouds and core shroud repair, the staff concurs with the applicant that the design of core shroud repair hardware for fatigue effects is valid for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(i).

In RAI 4.3.2.2, the staff requested that the applicant provide justification why vibration levels resulting from conditions such as increased core flow or single recirculation loop operation do not cause concerns for fatigue of jet pump riser braces or other internal components at Dresden Unit 3 or Quad Cities, similar to those at Dresden Unit 2. The applicant stated in its response that the potential for flow-induced vibration of reactor internals at Dresden and Quad Cities was evaluated as part of the EPU. The EPU analyses indicated that, except for the Dresden Unit 2 jet pump riser braces, the Dresden and Quad Cities plants can operate at the increased flow associated with EPU conditions for a 60-year plant life without exciting the safety-related reactor internal components above their established vibration limits during balanced (dual loop) recirculation flow operation and without developing resonance problems

4-28 because of vane passing frequency excitation. The EPU analyses also considered single recirculation loop operation and indicated that, with the existing flow restrictions that apply to single recirculation loop operation, there is no resonance problem because of vane passing frequency excitation at EPU operating conditions. The exception involving the Dresden Unit 2 jet pump riser braces occurs because these braces were designed differently from the Dresden Unit 3 and Quad Cities jet pump riser braces. In accordance with LRA Section 4.3.2.2, the applicant committed to repair or replace the Dresden Unit 2 jet pump riser braces before the period of extended operation. This is Commitment #48 in Appendix A of this SER.

The applicant also stated in LRA Section 4.3.2.2 that the repaired or replaced braces will be qualified for the period of extended operation, in accordance with the requirements of 10 CFR 54.21(c)(1)(ii). The staff finds this acceptable because it will provide a margin of safety for the Dresden Unit 2 braces similar to that of Dresden Unit 3 and Quad Cities.

The applicants supplements for the Dresden and Quad Cities UFSARs regarding the fatigue analyses of reactor vessel internals are provided in Section A.3.2.2 of the respective LRAs.

The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.3.2 of the LRA.

4.3.2.3 ASME Code,Section III, Class 1 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analysis The applicant has identified CLB ASME Code,Section III, Class 1 fatigue analyses associated with the recirculation piping system at Dresden Unit 3 as TLAAs, in conformance with the provisions of 10 CFR 54.21(c)(1) and the piping components listed in the appropriate tables in NUREG-1801. The applicant listed the 40-year CUFs for several components which exceeded a threshold value of 0.4. The CUFs of these components therefore have the potential of exceeding the limiting value of 1.0 during the period of extended operation. The staff reviewed these locations and determined that they conform with components listed in NUREG-1801. Pursuant to 10 CFR 54.21(c)(1)(iii), the applicant has therefore committed to monitor, as part of the Metal Fatigue AMP, seven components where the CUFs are expected to exceed the threshold value during the 40-year plant life. This is Commitment #34 in Appendix A of this SER. The applicant stated that as part of the Metal Fatigue AMP, all necessary plant transient events will be tracked and the CUFs calculated, to verify that the CUF remains less than 1.0 for all monitored components. In the event that the CUF for any component is projected to exceed 1.0 before 60 years of operation, the applicant commits to implement appropriate corrective actions in accordance with the Exelon Corrective Action Program, described in Section B.2.1 of the LRA, before the period of extended operation. This is Commitment #39 in Appendix A of this SER. The staff finds this acceptable because the applicant has provided assurance that an adequate margin of safety for the recirculation piping system at Dresden Unit 3 will be maintained for the period of extended operation, as reflected in its commitment to meet the ASME Code,Section III, Class 1 fatigue analysis criterion within the Metal Fatigue AMP or to implement corrective actions associated with the Exelon Corrective Action Program.

The applicants supplement for the Dresden UFSAR regarding the recirculation piping system fatigue analyses is provided in Section A.3.2.3.1 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.3.3.1 of the LRA.

4-29 4.3.2.4 Reactor Coolant Pressure Boundary Piping and Components Designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C The applicant indicated that RCPB piping systems other than the recirculation piping at Dresden Unit 3 were designed to the requirements of USAS B31.1, 1967 Edition, or ASME Code,Section III, Classes 2 and 3. These codes consider fatigue implicitly in the design calculations by applying stress range reduction factors to the allowable stress range to account for cyclic thermal conditions. The applicant approximated the number of cycles over a 40-year plant life by the thermal cycles used in the reactor vessel fatigue analysis. These thermal cycles are listed in UFSAR Table 3.9-1. For a 60-year plant life, the total count of all these listed thermal cycles is less than 3300. This is substantially less than the 7000 cycle limit in USAS B31.1 or ASME Code,Section III, Classes 2 and 3. The staff therefore finds that the applicant has demonstrated that an adequate margin of safety for the RCPB piping systems at Dresden Unit 2 and Quad Cities designed to USAS B31.1 or ASME Code,Section III, Classes 2 and 3 will be maintained for the period of extended operation, because the margin conforms with accepted industry practice. The staff concludes that the applicant has demonstrated that the existing analyses will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

The applicants supplement for the Dresden and Quad Cities UFSAR regarding the RCPB piping system fatigue analyses is provided in Section A.3.2.3.2 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.3.3.2 of the LRA.

4.3.2.5 Fatigue Analysis of the Isolation Condenser In Section B.1.34, the applicant stated that as an enhancement to the Metal Fatigue AMP the program will provide for tracking of fatigue stress cycles for the Dresden isolation condensers.

In RAI 4.3.3.3, the staff requested that the applicant provide an explanation of why this enhancement to the AMP does not conflict with the disposition of this item stated in Section 4.3.3.3 of the LRA. (The applicant indicated that the fatigue analysis remains valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).)

The applicant stated that the isolation condensers are included in the manual cycle counting option of the FatiguePro monitoring program of the Metal Fatigue AMP because RCPB piping locations associated with the condenser have predicted CUFs greater than 0.4. Because the thermal transients associated with the isolation condensers can affect the CUFs of these locations, the applicant has conservatively elected to include the isolation condenser locations in the monitoring program. In accordance with 10 CFR 54.21(c)(1)(iii), the staff finds this acceptable because cycle monitoring of the isolation condensers as proposed by the applicant will provide assurance that the margin of safety of the condensers and associated piping will be maintained for the period of extended operation, in conformance with the CLB design requirements of the condensers.

The applicants supplement for the Dresden UFSAR regarding the isolation condenser fatigue analysis is provided in Section A.3.2.3.3 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.3.3.3 of the LRA.

4-30 4.3.2.6 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)

GSI-166, Adequacy of the Fatigue Life of Metal Components, raised concerns regarding the conservatism of the fatigue curves used in the design of the RCS components. Although GSI-166 was resolved for the current 40-year design life of operating components, the staff identified GSI-190, Fatigue Evaluation of Metal Components for 60-Year Plant Life, to address license renewal. The NRC closed GSI-190 in December 1999, with the following conclusion:

The results of the probabilistic analyses, along with the sensitivity studies performed, the iterations with industry (NEI and EPRI), and the different approaches available to the licensees to manage the effects of aging, lead to the conclusion that no generic regulatory action is required, and that GSI-190 is closed. This conclusion is based primarily on the negligible calculated increases in core damage frequency in going from 40 to 60 year lives. However, the calculations supporting resolution of this issue, which included consideration of environmental effects, and the nature of age-related degradation indicate the potential for an increase in the frequency of pipe breaks as plants continue to operate. Thus, the staff concludes that, consistent with existing requirements in 10 CFR 54.21, licensees should address the effects of coolant environment on component fatigue life as aging management programs are formulated in support of license renewal.

The applicant has committed to evaluate the component locations listed in NUREG/CR-6260, that are applicable to an older vintage BWR plant for the effect of the environment on the fatigue life of the components, not including the RCIC tee located on the feedwater line. This is Commitment #48 in Appendix A of this SER. The applicant justified this on the basis that there is no RCIC system at Dresden and therefore this component does not exist at Dresden, and that at Quad Cities this tee is located in the Class 2 portion of the feedwater line.

However, the applicant stated that for both plants, alternate limiting Class 1 locations on the feedwater piping will be evaluated for environmental fatigue effects.

The applicant stated that for each location, detailed environmental fatigue calculations will be performed using the appropriate Fen relationships from NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, for carbon and alloy steels, and those from NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, for stainless steel, as appropriate for the material. These calculations will be completed prior to the period of extended operation, and appropriate corrective action will be taken if the resulting end-of-life CUF values exceed 1.0. The staff finds this acceptable because it concurs with the staff position on environmental effects on metal fatigue of Class 1 components.

The applicant also stated that it reserves the right to modify this position in the future, based on the results of ongoing industry activities on this topic, subject to NRC approval prior to changes in this position. The staff concurs with this statement since any licensee always has the option of submitting a license amendment that would be subject to review and approval by the staff.

In RAI 4.3.4, the staff requested that the applicant identify the alternate limiting Class 1 feedwater piping locations and provide the calculated CUF for these locations. In response, the applicant stated that the entire Class 1 portion of a feedwater line was structurally modeled, based on the bounding geometry for all eight feedwater loops (two loops per unit),

4-31 and a fatigue analysis was performed using ASME Code,Section III, NB-3600 methodology.

The results of the fatigue analysis determined that the location with the highest fatigue usage (CUF=0.0859, not including environmental effects) was at the tee joining a riser pipe to a header (Node 15a in the structural model). This location will be used to perform plant-specific environmental fatigue calculations. (This location is different from the reactor vessel feedwater nozzle location specified in NUREG/CR-6260, since this location will be separately evaluated for environmental effects.) The staff finds this acceptable because the selection of the limiting location was performed in accordance with the basis for selection of the locations for older vintage BWR plants, stated in NUREG/CR-6260 for a similar portion of the feedwater lines.

In accordance with 10 CFR 54.21(d), the applicant has included a section addressing the effects of reactor coolant environment on fatigue life of components and piping (Issue 190) in the UFSAR Supplement Section A.3.2.4 for Dresden and for Quad Cities. The applicant has committed to perform plant-specific calculations for environmental effects on the fatigue life of the components listed in NUREG/CR-6260 and the limiting location on the feedwater line, and take appropriate corrective actions if the resulting projected end-of-life CUF values exceed 1.0.

This is Commitment #48 in Appendix A of this SER. The calculations will include appropriate environmental fatigue effect factors from NUREG/CR-6583 and NUREG/CR-5704. The applicant also stated in the UFSAR Supplement that it reserves the right to modify this position in the future, based on the results of ongoing industry activities on this topic, subject to NRC approval prior to changes in this position. The staff concurs with this statement since any licensee always has the option of submitting a license amendment that would be subject to review and approval by the staff. The staff finds this supplement acceptable because it provides a reasonable summary of the information presented in Section 4.3.4 of the LRA.

4.3.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i), that the analyses remain valid for the period of extended operation for the low-cycle thermal fatigue analysis of the core shroud and repair hardware; the reactor coolant pressure boundary piping and components designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C; and the fatigue analysis of the isolation condenser TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description, as required by 10 CFR 54.21(d), of the low-cycle thermal fatigue analysis of the core shroud and repair hardware; the reactor coolant pressure boundary piping and components designed to USAS B31.1, ASME Code,Section III, Classes 2 and 3, or ASME Code,Section VIII, Classes B and C; and the fatigue analysis of the isolation condenser TLAA evaluations for the period of extended operation.

On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(ii), that the analyses have been projected to the end of the period of extended operation for the high-cycle flow-induced vibration fatigue analysis of jet pump riser braces and the effects of reactor coolant environment on fatigue life of components and piping TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the high-cycle flow-induced vibration fatigue analysis of jet pump riser braces and the effects of reactor coolant environment on fatigue life

4-32 of components and piping TLAA evaluations for the period of extended operation, as reflected in the license condition required by 10 CFR 54.21(d).

On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(iii), that the effects of aging on the intended functions will be adequately managed for the period of extended operation for the reactor vessel fatigue analyses and the ASME Code,Section III, Class 1 reactor coolant pressure boundary piping and component fatigue analysis TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the reactor vessel fatigue analyses and the ASME Code,Section III, Class 1 reactor coolant pressure boundary piping and component fatigue analysis TLAA evaluations for the period of extended operation, as required by 10 CFR 54.21(d).

Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.4 Environmental Qualification 4.4.1 Environmental Qualification Program TLAA The NRC has established nuclear station environmental qualification (EQ) requirements in Crieterion 4 of Appendiz A to 10 CFR Par 50 and 10 CFR 50.49. The latter specifically requires that an EQ Program be established to demonstrate that certain electrical components located in harsh plant environments (that is, those areas of the plant that could be subject to the harsh environmental effects of a LOCA, high-energy line breaks (HELBs), or post-LOCA radiation) are qualified to perform their safety function in those harsh environments after the effects of inservice aging. Also, 10 CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of EQ. For the purpose of license renewal, only those components with a qualified life of 40 years or greater would require TLAAs.

The staff has reviewed LRA Section 4.4, Environmental Qualification of Electrical Equipment in which the applicant described the technical bases and justification for why the Dresden and Quad Cities EQ Programs, together with other plant programs and processes, adequately manages the effects of aging on the intended function(s) of electrical components for the period of extended operation. The staff reviewed this section of the LRA to determine whether the applicant had demonstrated that the effects of aging on the intended function(s) of the electrical equipment will be adequately managed through the Dresden and Quad Cities EQ Programs, together with other programs and processes, during the period of extended operations, as required by 10 CFR 54.21(c)(1)(iii).

4.4.1.1 Summary of Technical Information in the Application The Dresden and Quad Cities EQ Programs are established to demonstrate that certain electrical components located in an environment that is subject to a LOCA, HELB, or post-LOCA radiation are qualified to perform their safety function after inservice aging. The Dresden and Quad Cities EQ Programs comply with the requirements of 10 CFR 50.49(e)(5) for aging considerations that affect functionality and make provisions to replace the

4-33 components or establish ongoing qualification when the demonstrated qualified life has expired. The EQ-related equipment is identified in a controlled equipment database with a qualification binder that is maintained with records on performance specifications, electrical characteristics and environmental conditions.

The Dresden and Quad Cities EQ Programs manage thermal, radiation and cyclic aging as applicable for all electrical components within the scope of 10 CFR 50.49 and for components that are presently qualified in accordance with the DOR guidelines. Compliance with 10 CFR 50.49 provides evidence that the component will perform its intended functions during and after a design-basis event after experiencing the effects of inservice aging.

Under 10 CFR 54.21(c)(1)(iii), the Dresden and Quad Cities EQ Programs, which implement the requirements of 10 CFR 50.49 (as further clarified by the DOR guidelines, NUREG-0588, and RG 1.89, Revision 1), are viewed as an AMP for license renewal. Aging evaluations of electrical components in the Dresden and Quad Cities EQ Programs that specify qualification of at least 40 years are TLAAs. Reanalysis will be applied to EQ components now qualified for the current operating term of 40 years.

The reanalysis of an aging evaluation may be performed to extend the qualification by reducing margin or excess conservatism incorporated in the prior evaluation. Reanalysis of an aging evaluation to extend the qualification of a component may be performed as part of the EQ Program. While a component life-limiting condition may be result from thermal, radiation, or cyclical aging, the vast majority of component aging limits are based on thermal conditions.

Conservatism may exist in aging evaluation parameters, such as the assumed ambient temperature of the component, unrealistically low activation energy, or in the application of a component as de-energized instead of energized. The important attributes of reanalysis will include analytical methods, data collection and conservative reduction methods, underlying assumptions, acceptance criteria, and corrective actions (if acceptance criteria are not met), as discussed below.

Analytical Methods: The analytical models used in the reanalysis of an aging evaluation are the same as those previously applied during the previous evaluation. The Arrhenius methodology is an acceptable thermal model for performing an aging evaluation. The analytical method used for a radiation aging evaluation demonstrates qualification for the total integrated dose (that is, normal radiation dose for the projected installed life plus accident radiation dose). For license renewal, one acceptable method of establishing the 60-year normal radiation dose is to multiply the 40-year normal radiation dose by 1.5 (that is, 60 years/40 years). The result is added to the accident radiation dose to obtain the total integrated dose for the component. For cyclical aging, a similar approach may be used. Other models may be justified on a case-by-case basis.

Data Collection and Reduction Methods: Reducing excess conservatism in the component service conditions (for example, temperature, radiation, and cycles) used in the previous aging evaluation is the chief method used for a reanalysis. Temperature data used in an aging evaluation should be conservative and based on plant-design temperatures or on actual plant temperature data. When used, plant temperature data can be obtained in several ways, including monitors used for technical specification compliance, other installed monitors, measurements made by plant operators during rounds, and temperature sensors on large motors (while the motor is not running). When used, a representative number of temperature

4-34 measurements is conservatively evaluated to establish the temperatures used in an aging evaluation. Plant temperature data may be used in an aging evaluation in different ways, such as (1) directly applying the plant temperature data in the evaluation, or (2) using the plant temperature data to demonstrate conservatism when using plant-design temperatures for an evaluation. Any changes to material activation energy values as part of a reanalysis are justified on a case-specific basis. Similar methods of reducing excess conservatism in the component service conditions used in prior aging evaluations may be used for radiation and cyclical aging.

Underlying Assumptions: EQ component aging evaluations contain sufficient conservatism to account for most environmental changes occurring because of plant modifications and events.

When unexpected adverse conditions are identified during operational or maintenance activities that affect the normal operating environment of a qualified component, the affected EQ component is evaluated and appropriate corrective actions are taken, which may include changes to the qualification bases and conclusions.

Acceptance Criteria and Corrective Actions: The reanalysis of an aging evaluation could extend the qualification of the component. If the qualification cannot be extended by reanalysis, the component is maintained, replaced, or requalified prior to exceeding the period for which the current qualification remains valid. A reanalysis is performed in a timely manner (that is, sufficient time is available to maintain, replace, or requalify the component if the reanalysis is unsuccessful).

Environmental Qualification of Electrical Components: As stated in LRA Section B.1.35, the Environmental Qualification of Electrical Components AMP is implemented through station procedures and predefined tasks. The Dresden and Quad Cities EQ Programs comply with 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants. All EQ equipment is included within the scope of license renewal. The program provides for maintenance of the qualified life for electrical equipment important to safety within the scope of 10 CFR 50.49. Program activities establish, demonstrate, and document the level of qualification, qualified configuration, maintenance, surveillance, and replacement requirements necessary to meet 10 CFR 50.49. Qualified life is determined for equipment within the scope of the EQ Program, and appropriate actions such as replacement or refurbishment are taken prior to or at the end of the qualified life of the equipment so that the aging limit is not exceeded.

NUREG-1801 Consistency: The Environmental Qualification of Electrical Components AMP is consistent with the 10 elements of AMP X.E1, Environmental Qualification (EQ) of Electrical Components, specified in NUREG-1801.

Operating Experience: The Environmental Qualification of Electrical Components AMP provides for consideration of operating experience to reconcile qualification bases and conclusions, including the equipment qualified life. Operating experience and information related to systems, equipment, or components, as reported through NRC bulletins, notices, circulars, generic letters, and Part 21 notifications, are evaluated for applicability. The evaluations are documented and corrective actions are identified. Operating experience has demonstrated that the program manages aging as required by 10 CFR 50.49. When problems have been identified through industry or plant-specific experience, corrective actions have been taken to prevent recurrence.

4-35 4.4.1.2 Staff Evaluation The staff reviewed the information in Section 4.4 of the LRA to determine whether the applicant has demonstrated that the effects of aging on the intended function(s) of electrical components will be adequately managed through its existing program, together with other plant programs/processes, during the period of extended operation as required by 10 CFR 54.21(c)(1)(iii).

The applicants program activities establish, demonstrate, and document the level of qualification, qualified configuration, maintenance, surveillance, and replacement requirements necessary to meet 10 CFR 50.49. Qualified life is determined for equipment within the scope of the EQ Program and appropriate actions, replacement, or refurbishment are taken prior to or at the end of qualified life of the equipment so that aging limits or acceptable margins are not exceeded. In response to RAI 4.3-0, the applicant committed to the following (Commitment #49 in Appendix A of this SER) in a letter dated October 3, 2003:

The Dresden/Quad Cities license renewal evaluations were based upon the plant environmental conditions associated with EPU implementation. Prior to the period of extended operation, the Environmental Qualification (EQ) Binders for components within the scope of 10 CFR 50.49 will be updated to include environmental conditions associated with EPU implementation together with an extended operating period of 60 years.

The applicants Environmental Qualification of Electrical Components AMP provides reasonable assurance that aging effects are adequately managed so that the intended functions of components within the scope of 10 CFR 50.49 are maintained during the period of extended operation.

The staff also reviewed the UFSAR Supplement for this TLAA and concluded that it provided an adequate summary description of the TLAA to satisfy the requirement of 10 CFR 54.21(d).

4.4.1.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(iii), that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation for the EQ of electrical equipment TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the EQ of electrical equipment TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.4.2 Generic Safety Issue 168, Environmental Qualification of Low-Voltage Instrumentation and Control (I&C) Cables 4.4.2.1 Summary of Technical Information in the Application The applicant states that NRC guidance for addressing GSI-168 Environmental Qualification of Low Voltage Instrumentation and Control (I&C) Cables, for license renewal is contained in the June 2, 1998, NRC letter to NEI. In the letter, the NRC states:

4-36 With respect to addressing GSI-168 for license renewal, until completion of an ongoing research program and staff evaluations the potential issues associated with GSI-168 and their scope have not been defined to the point that a license renewal applicant can reasonably be expected to address them at this time. Therefore, an acceptable approach described in the Statement of Consideration is to provide a technical rationale demonstrating that the current licensing basis for environmental qualification pursuant to 10 CFR 50.49 will be maintained in the period of extended operation. Although the Statement of Consideration also indicated that an applicant should provide a brief description of one or more reasonable options that would be available to adequately manage the effects of aging, the staff does not expect an applicant to provide the options at this time.

This is consistent with the above NRC guidance, no additional information is required to address GSI-168 in a license renewal application at this time.

4.4.2.2 Staff Evaluation GSI-168, Environmental Qualification of Low Voltage Instrumentation and Control (I&C)

Cables, was developed to address environmental qualification of electrical equipment. The staff guidance to the industry (letter dated June 2, 1998, from Mr. C.I. Grimes (NRC) to Mr.

D,J. Walters (NEI) states the following:

GSI-168 issues have not been identified to a point that a license renewal applicant can be reasonably expected to address these issues, specifically at this time An acceptable approach is to provide a technical rationale demonstrating that the CLB for EQ will be maintained in the period of extended operation.

For the purpose of license renewal, as discussed in the Statements of Consideration (60 FR 22484, May 8, 1995), there are three options for addressing issues associated with a GSI:

If the issue is resolved before the renewal application is submitted, the applicant can incorporate the resolution in the LRA.

An applicant can submit a technical rationale that demonstrates that the CLB will be maintained until some later point in the period of extended operation, at which time one or more reasonable options would be available to adequately manage the effects of aging.

An applicant can develop a plant-specific AMP that incorporates the resolution of the aging issue.

On May 2, 2003, the staff issued NRC Regulatory Issue Summary (RIS) 2003-09, Environmental Qualification of Low-Voltage Instrumentation and Control Cables, providing the results of the staffs technical assessment of GSI-168 following completion of the NRC-sponsored cable test research. The staff concluded that typical I&C cable qualification test programs include numerous conservative practices that collectively provide a high level of confidence that the installed I&C cables will perform their intended functions during and following design-basis events, as required by 10 CFR 50.49. These conservative practices continue to support the current use of a single prototype test specimen during qualification testing, and as such, a successful test provides a high level of confidence that these cables will be able to perform their safety functions during and following design-basis events. However, I&C cable LOCA test failures during the NRC-sponsored research program indicate that the

4-37 original margin and conservatism inherent in the qualification process have been reduced.

Therefore, licensee awareness of the operating service environments (temperature, radiation, and moisture) is essential to ensure that the operating conditions in nuclear power plants do not exceed the qualification parameters that were assumed during qualification testing.

RIS 2003-09 further states that licensees that have addressed license renewal recognize that knowledge of the operating service environments is essential to extending the qualified life of I&C cables. Where measured environmental service conditions are less severe than those used in the original qualification and when the cables are not degraded, the licensee assessed the difference between the operating environment and the original qualification environment to extend the qualified life of the cables to 60 years by reanalysis. This approach, based on the Arrhenius methodology, has been found acceptable by the staff during its review of license renewal applications.

In response to RAI 4.4-1, the applicant provided the following in a letter dated October 3, 2003:

Exelon performed an analysis for all EQ-related equipment and has qualified all low-voltage I&C cables for 60 years of service without lowering the original environmental service conditions. No measured environmental service conditions were used in the analysis, but the qualification files have not been updated to support 60-year life. For Dresden and Quad Cities, the EQ TLAA ensures the effects of aging will be adequately managed for the period of extended operation per the provision of 10 CFR 54.21(c)(1)(iii). Therefore, with respect to GSI-168, Section 4.4 of the application should have stated: Adherence to the EQ program and use of current EQ process for Dresden and Quad Cities stations will provide reasonable assurance through the extended period of operation that the equipment qualification will be maintained in compliance with the applicable NRC requirements.

The staff finds that the applicant has adequately addressed the issues associated with GSI-168 for license renewal because the applicant will continue to manage the effects of aging through the EQ Program for the period of extended operation.

4.4.2.3 Conclusions On the basis of its review, the staff concludes that the applicant has adequately addressed the issues associated with GSI-168 for license renewal as required by 10 CFR 54.21(c)(1)(iii). The staff has also reviewed the UFSAR Supplement and the staff concludes the applicant has provided an adequate description of its evaluation of this TLAA for the period of extended operation as required by 10 CFR 54.21(d).

4.5 Loss of Prestress in Concrete Containment Tendons None of the Dresden or Quad Cities containments have prestress tendons. As such, this topic is not a TLAA.

4.6 Fatigue of Primary Containment, Attached Piping, and Components The applicant stated in Section 4.6 of the LRA that the cyclic loads acting on the primary containment and the attached piping and components include reactor building interior temperature variation during the heatup and cooldown of the RCS, a LOCA, annual outdoor temperature variations, thermal loads on containment penetrations because of high-energy

4-38 piping lines (such as steam and feedwater lines), seismic loads, and pressurization resulting from periodic Type A integrated leak rate tests.

The metal containments, penetration sleeves (including dissimilar metal welds), and penetration bellows may be designed in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. If a plants code of record requires a fatigue analysis, then this analysis may be a TLAA and must be evaluated in accordance with 10 CFR 54.21(c)(1) to ensure that the effects of aging on the intended functions will be adequately managed for the period of extended operation.

The adequacy of the metal containment, penetration sleeves, and penetration bellows is reviewed for the period of extended operation. Section 4.3 of this SER includes a separate review of the fatigue analysis of the pressure boundary of process piping.

The primary containments at Dresden and Quad Cities were designed in accordance with ASME Code Section III, 1965 Edition with Addenda up to and including Winter 1965.

Subsequent to the original design, elements of the Dresden and Quad Cities containments were reanalyzed in response to discoveries of unevaluated loads because of assumed pressure and temperature cycles resulting from SRV discharge and design-basis LOCA events. This reevaluation was performed under Mark I Containment Program plant-unique analyses and reported in plant unique analysis reports (PUARs).

The applicant also stated the following:

In the absence of hydrodynamic loads, fatigue is not a concern in containment design except at penetrations or other stress concentration areas. The licensing and design basis documents do not reflect the existence of any fatigue analyses for the drywell or its penetrations. However, the drywell process bellows, including replacement bellows for Quad Cities, were specified for a finite number of operating cycles.

4.6.1 Summary of Technical Information in the Application 4.6.1.1 Fatigue Analysis of the Suppression Chamber, Vents and Downcomers The applicant stated that the Dresden and Quad Cities PUARs describe the fatigue analyses of the suppression chambers and suppression chamber (torus) vents, including the vent headers and downcomers. The analyses assumed a limited number of SRV actuations during the plant transients, based on a survey of plant data extrapolated to 40 years. Each SRV actuation was assumed to result in one thermal, one pressure, and five dynamic load cycles.

The applicant listed the transients and corresponding design cycles used for the fatigue analysis of the suppression chamber shells and the associated welds, as well as the suppression chamber vent headers, downcomers, and associated welds for both Dresden and Quad Cities. Table 4.6-1 of the LRA lists the calculated CUFs exceeding 0.4 for a 40-year period of operation for each component and associated weld. The worst-location CUF is 0.92, located at the intersection of the vent headers and the downcomers.

For most shell, vent, and penetration locations, the predicted 40-year CUF is less than 0.666.

In accordance with 10 CFR 54.21(c)(1)(i), the fatigue analyses of these locations remain valid

4-39 because the CUFs for the period of extended operation will be less than 1.0. However, the applicant stated that a CUF of 0.666 provides no analytical or event margin. The applicant stated that the validation will therefore be applied to locations with calculated 40-year CUFs less than 0.4. The locations where this threshold value is expected to be exceeded will be included in the Metal Fatigue AMP, in accordance with 10 CFR 54.21(c)(1)(iii). These locations, listed in Table 4.6.1-2 of the LRA, consist of the suppression chamber weld and the vent header at the downcomer-vent header intersection (for Dresden Units 2 and 3) and the suppression chamber shell (for Quad Cities Units 1 and 2). These locations represent or bound all other locations for both plants. All necessary plant transients will be tracked to ensure that the CUF remains less than 1.0 for all monitored components. In the event fatigue usage is predicted to exceed 1.0 for any component before 60 years of operation, appropriate action will be taken in accordance with the Exelon Corrective Action Program, described in Section B.2.1 of the LRA. The required implementing actions will be completed prior to the period of extended operation.

4.6.1.2 Fatigue Analysis of Safety Relief Valve Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations The applicant stated that the PUARs for each site describe the fatigue analyses of the SRV discharge lines and their penetrations through the vent lines, the suppression chamber (torus) shell and the attached piping systems, and the T-quenchers inside the suppression chamber.

These analyses assume a limited number of SRV actuations throughout the 40-year life of the plant and are therefore TLAAs.

The fatigue analysis of the external torus attached piping and Class 2/3 SRV discharge lines assumed 800 SRV actuations with 5 cycles per actuation for a 40-year plant life. Other thermal cycles resulting from normal operating conditions were considered to have a negligible effect.

The analysis indicated that these lines would have a CUF less than 0.5 at the end of 40 years of operation.

The SRV discharge line-vent line penetrations and associated sections of the SRV discharge lines are classified as Class MC components. The fatigue analyses of these components were based on 110 isolation events, each with 2 actuations. The analyses show maximum CUFs of 0.09 for Dresden and 0.18 for Quad Cities.

The fatigue calculations of the Class MC penetration components and attachments assumed 220 SRV actuations of 5 dynamic cycles each, plus 4050 cycles caused by condensation oscillation or chugging. The alternating stress intensity corresponding to the total sum of these cycles was determined. These analyses did not calculate the CUFs. Instead, the applicant specified that the alternating stress intensity at the highest stress location should be smaller than that corresponding to the assumed number of cycles. This, in effect, ensures that the CUF will be less than 1.0 for the entire period of plant operation. This concept is applicable only with the stress monitoring option of the Metal Fatigue AMP, evaluated in Section 3.1.2.3.9 of this SER.

To address concerns associated with potential plugging and unacceptable head loss, the emergency core cooling system (ECCS) suction strainers were replaced with larger units at both Dresden and Quad Cities. This replacement required revised stress and fatigue analyses.

The fatigue analyses of these penetrations were based on the generic Mark I owners group

4-40 analysis for 40-years of plant life. For Dresden Unit 2, the highest calculated CUF for these components is 0.142. For Dresden Unit 3, the highest CUF for these components is 0.0832.

For Quad Cities, the highest calculated CUF for these components is 0.3087.

The highest 40-year CUFs for the suppression pool shell attached piping, SRV discharge lines, and penetrations are listed in Table 4.6.2-1 of the LRA. The applicant stated that th Metal Fatigue AMP, described in Section B.1.34 of the LRA, will include all locations where the 40-year CUF exceeds the threshold value of 0.4. Since only the SRV load cases contribute to fatigue of these components during normal operation, the contribution to the CUF at a particular location resulting from the number of SRV actuations will be monitored to ensure that it does not exceed 1.0 minus the CUF contribution from a postulated LOCA plus operating based events.

4.6.1.3 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses The applicant stated that Mark I containment designs include a drywell-to-suppression chamber vent line. A bellows assembly is provided at the penetration of the vent line to the suppression chamber. The bellows allows differential movement of the vent system and suppression chamber without developing significant interaction loads. The fatigue analyses of these bellows are included in the Mark I Containment Long-Term Program plant-unique analysis.

Dresden and Quad Cities have external vacuum breakers which include 24-in. bellows. There are differences between the bellows configurations at the two plants, but the effect of these differences in the overall vent system analysis was found to be insignificant. At Quad Cities, these bellows were included in the vent system analysis. At Dresden, these bellows were evaluated as part of the penetrations. The applicant referenced an NRC SER which concluded that the owners comparison of the Dresden external vacuum breakers to the stress criteria of ASME Code,Section III, Subsection NC, 1977 Edition with Summer 1977 Addenda, demonstrated that the fatigue usage factors in the vacuum breakers and their attachments (and at Dresden, their penetrations) are bounded by other analyzed locations and therefore are not limiting. The SER therefore concluded that the design was adequate.

The applicant stated that the analysis of the vent line bellows was based on a total of 150 thermal and internal pressure-load cycles for the 40-year life of the plant and are therefore adequate for fatigue, since they have a rated capacity of 1000 cycles at maximum displacement. The number of load cycles for the period of extended operation is 225, which is less than 25 percent of the rated capacity. The applicant indicated that, pursuant to 10 CFR 54.21(c)(1)(i), the fatigue adequacy of the bellows is valid for the period of extended operation.

4.6.1.4 Primary Containment Process Penetration Bellows Fatigue Analysis The applicant stated in Section 4.6.4 of the LRA that at Dresden and Quad Cities the only containment process piping expansion joints subject to significant thermal expansion and contraction are those between the drywell shell penetrations and process piping. Thermal cycles on the bellows are imposed by thermal cycles experienced by the attached piping. The assumed thermal-cycle count can be approximated by the thermal cycles used in the reactor vessel fatigue analysis, which are listed in Table 3.9-1 of the UFSAR. The total count of all these cycles is 2200 for a 40-year plant life. For a 60-year plant life, the number of thermal

4-41 cycles for the piping penetration bellows analyses increases to 3300. The bellows are designed for 7000 operating cycles, corresponding to the design code for the piping. The applicant indicated that, pursuant to 10 CFR 54.21(c)(1)(i), the containment penetration bellows fatigue analysis remains valid for the period of extended operation.

4.6.2 Staff Evaluation The staff has reviewed the technical information in Section 4.6 of the LRA regarding fatigue of primary containment, attached piping, and components. In RAI 4.6, the staff requested justification for the statement that in the absence of hydrodynamic loads, fatigue is not a concern in containment design, except at penetrations or other stress concentration areas. The applicant stated that the primary containments for Dresden and Quad Cities were initially designed in accordance with ASME Code,Section III, 1965 Edition, with Addenda up to and including Winter 1965. These containments are Mark I containments that were originally designed to stress limits without requiring fatigue analyses (Class B vessels). The discovery of significant hydrodynamic loads caused by SRV discharges and small, intermediate and design basis pipe break discharges into the torus suppression pool, for which the containment had not been initially analyzed, was identified as an unresolved safety issue by the NRC, requiring the reanalysis of the torus, vents, and torus attached piping and internal structures, including fatigue analysis at limiting locations. The generic suppression pool hydrodynamic load definition and structural assessment techniques that were to be used to design plant modifications necessary to restore the Mark I containments were described in NUREG-0661, Safety Evaluation ReportMark I Containment Long Term Program. This report specified a long-term program to establish the structural and mechanical elements that were to be analyzed in a plant unique analysis. The long-term program plant-unique analyses for Dresden and Quad Cities were reported in a PUAR for each plant, which demonstrated that all applicable Mark I criteria in NUREG-0661 had been met. The NRC staff performed a post-implementation audit review of the PUARs and concluded that with the required containment modifications, the original design margins of the containment had been restored. The staff has reviewed this response and concludes that the applicant has provided satisfactory justification for the statement questioned in RAI 4.6.

The applicants supplement for the Dresden and Quad Cities UFSARs regarding containment fatigue is provided in Section A.3.4 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.6 of the LRA.

4.6.2.1 Fatigue Analysis of the Suppression Chamber, Vents, and Downcomers The staff has reviewed the technical information in Section 4.6.1 of the LRA regarding the fatigue TLAA of the suppression chamber, vents, and downcomers. The staff has also reviewed the applicants disposition of this TLAA and finds it acceptable because the applicant has selected a threshold limit of CUF=0.4 for 40 years of operation as a criterion for determining if the fatigue analyses performed under the Dresden and Quad Cities PUARs will remain valid for the period of extended operation. The staff concurs with the applicant that this criterion will provide additional analytical or event margin over the minimum CUF value of 0.666 for the period of extended operation. Those locations not exceeding the threshold criterion will therefore remain valid for the period of extended operation, in accordance with 10 CFR

4-42 54.21(c)(1)(i). In accordance with 10 CFR 54.21(c)(1)(iii), for locations where the CUF exceeds the criterion above, the staff finds the applicants commitment to manage the effects of fatigue for the period of extended operation with the Metal Fatigue AMP (B.1.34) acceptable because it will provide assurance that the monitored CUF at a location will not exceed the ASME Code,Section III CUF limit of 1.0. If the CUF is projected to exceed this limit the applicant has committed to take appropriate corrective action, as stated in Section 4.6.1 of the LRA, in accordance with the Exelon Corrective Action Program (described in Section B.2.1 of the LRA).

This is Commitment #39 in Appendix A of this SER.

The applicants supplement for the Dresden and Quad Cities UFSARs regarding the suppression chamber, vents, and downcomers fatigue TLAA is provided in Section A.3.4.1 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.6.1 of the LRA.

4.6.2.2 Fatigue Analysis of Safety Relief Valve Discharge Piping Inside the Suppression Chamber, External Suppression Chamber Attached Piping, and Associated Penetrations The staff has reviewed the technical information in Section 4.6.2 of the LRA regarding the fatigue TLAA of the SRV discharge piping inside the suppression chamber, the external suppression chamber attached piping, and the associated piping penetrations. The staff has also reviewed the applicants disposition of this TLAA and finds it acceptable because the applicant has selected a threshold limit of CUF=0.4 for 40 years of operation as a criterion for determining if the fatigue analyses performed under the Dresden and Quad Cities PUARs will remain valid for the period of extended operation. The staff concurs with the applicant that this criterion will provide additional analytical or event margin over the minimum CUF value of 0.666 for the period of extended operation. Those locations on the inside or outside of the piping not exceeding the threshold criterion will therefore remain valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i). In accordance with 10 CFR 54.21(c)(1)(iii), for locations where the CUF is projected to exceed the criterion above, the staff finds the applicants commitment to manage the effects of fatigue for the period of extended operation with the Metal Fatigue AMP (B.1.34) acceptable because the AMP will provide assurance that the monitored CUF at a location will not exceed the ASME Code,Section III CUF limit of 1.0. This is Commitment #34 of Appendix A of this SER. The staff also finds acceptable that the applicant will use the cycle counting capability of the Metal Fatigue AMP to monitor the number of SRV lifts for the SRV discharge piping penetration components and welds. This will implicitly assure that the CUF of these components will remain less than 1.0 for the period of extended operation.

In RAI 4.6.2, the staff requested that the applicant state what corrective action will be taken if the SRV lifts exceed the number required to ensure that the CUF at a location remains less than 1.0. In its response, the applicant stated that a condition report is required to be generated when the number of cycles approaches or exceeds the design allowable number.

This procedure, along with site implementing procedures, will be revised to ensure that a condition report will be generated before exceeding any design allowable cycle limit. This will allow appropriate corrective or mitigating actions to be taken before the number of SRV lifts exceeds the number required to ensure that the CUF remains less than 1.0. Corrective actions could include reanalysis of the CUF with more refined techniques or plant modifications as

4-43 appropriate. The staff finds the response acceptable because it conforms to current industry practice.

The applicants supplement for the Dresden and Quad Cities UFSARs regarding the fatigue TLAA of the SRV discharge piping inside the suppression chamber, the external suppression chamber attached piping, and the associated penetrations is provided in Section A.3.4.2 of the LRA. The staff has reviewed this supplement and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.6.2 of the LRA.

4.6.2.3 Drywell-to-Suppression Chamber Vent Line Bellows Fatigue Analyses The staff has reviewed the technical information in Section 4.6.3 of the LRA regarding the fatigue TLAA of the drywell-to-suppression chamber vent line bellows. The staff has also reviewed the applicants disposition of this TLAA and finds it acceptable because the applicant has demonstrated that in accordance with 10 CFR 54.21(c)(1)(i), the rated cyclic capacity of the drywell-to-suppression chamber vent line bellows is adequate for the number of pressure and temperature cycles expected during the period of extended operation.

The applicants supplement for the Dresden and Quad Cities UFSARs regarding the fatigue TLAA of the drywell-to-suppression chamber vent line bellows is provided in Section A.3.4.3 of the LRA. The staff has reviewed this supplement and finds it acceptable because it reflects the information presented in Section 4.6.3 of the LRA.

4.6.2.4 Primary Containment Process Penetration Bellows Fatigue Analysis The staff has reviewed the technical information in Section 4.6.4 of the LRA regarding the fatigue TLAA of the primary containment process penetration bellows. The staff has also reviewed the applicants disposition of this TLAA and finds it acceptable because the applicant has demonstrated that in accordance with 10 CFR 54.21(c)(1)(i), the rated cyclic capacity of the drywell-to-shell vent line bellows is adequate for the number of thermal cycles expected during the period of extended operation.

The applicants supplement for the Dresden and Quad Cities UFSARs regarding the fatigue TLAA of the primary containment process penetration bellows is provided in Section A.3.4.3 of the LRA. The staff has reviewed the supplemental section and finds it acceptable because it provides a reasonable summary of the information presented in Section 4.6.3 of the LRA.

4.6.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) and 10 CFR 54.21(c)(1)(iii), that the analyses remain valid for the period of extended operation. Further, the effects of aging on the intended functions will be adequately managed for the period of extended operation for the fatigue analysis of the suppression chamber, vents, and downcomers and the fatigue analysis of safety relief valve discharge piping inside the suppression chamber, external suppression chamber attached piping, and associated penetrations TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the fatigue analysis of the suppression chamber, vents, and downcomers and the fatigue analysis of safety relief valve

4-44 discharge piping inside the suppression chamber, external suppression chamber attached piping, and associated penetrations TLAA evaluations for the period of extended operation, as required by 10 CFR 54.21(d).

On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i), that the analyses remain valid for the period of extended operation for the drywell-to-suppression chamber vent line bellows and the primary containment process penetration bellows fatigue analysis TLAAs. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the fatigue analysis of the drywell-to-suppression chamber vent line bellows fatigue analyses and the primary containment process penetration bellows fatigue analysis TLAA evaluations for the period of extended operation, as required by 10 CFR 54.21(d).

Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7 Other Plant-Specific Time-Limited Aging Analyses 4.7.1 Reactor Building Crane Load Cycles 4.7.1.1 Summary of Technical Information in the Application The applicant stated that the reactor building overhead cranes at Dresden and Quad Cities were designed to meet the design fatigue loading requirements of the Crane Manufacturers Association of America (CMAA) Specification 70, Class A1. This evaluation of expected cycles over a 40-year plant life is the basis of a safety determination and is therefore a TLAA.

The applicant stated that the reactor building overhead crane is designed for 100,000 loading cycles. The weldments are categories B and C, which permit a stress range of 28,000-33,000 psi. The maximum allowable stress in the girders with rated load is 17,600 psi and the minimum stress (no load) is approximately 2,400 psi. The maximum stress range in the girders will not exceed 15,200 psi. Because the maximum stress permitted in other weldments is 14,000 psi, they have a smaller range and better fatigue resistance than the girders. According to the applicant these ranges are satisfactory for approximately 2 million loading cycles. This would be equivalent to approximately 50,000 125-ton loads per year handled in the center of the span over a 40-year period.

The applicant estimated that these cranes will see fewer than 5000 cycles at rated capacity and a larger number of cycles at significantly less than rated capacity. For this reason, the applicant indicated that fatigue life is not significant to the operation of this equipment.

4.7.1.2 Staff Evaluation The reactor building cranes are designed to CMAA-70 Class A1. Based on its review of the design evaluations the staff concurs with the applicant that all components are qualified for 100,000 loading cycles (i.e., 100,000 lifts at rated capacity). The staff also concurs with the applicants estimate of the maximum stress ranges of 15,200 psi and 14,000 psi for the reactor

4-45 building overhead cranes and the weldments respectively The 40-year estimated cycles equal most 5000 rated-capacity load cycles, or up to 7500 if extended to a 60-year life. The 60-year, 7,500-cycle estimate remains a small fraction of the 100,000 cycle minimum design. Therefore, fatigue life is not significant to the operation of this equipment and remains valid for the period of extended operation. The applicant has provided a satisfactory validation of 10 CFR 54.21(c)(1)(i).

4.7.1.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) that the analyses remain valid for the reactor building crane load cycles TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the reactor building crane load cycles TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.2 Metal Corrosion 4.7.2.1 Corrosion Allowance for Power-Operated Relief Valves 4.7.2.1.1 Summary of Technical Information in the Application The applicant stated in Section 4.7.2.1 of the LRA, Corrosion Allowance for Power Operated Relief Valves, that power-operated relief valves (PORVs) were installed at Quad Cities in 1995 to replace the main steam line electromagnetic relief valves. The specification for the PORVs is cited in the Quad Cities UFSAR Section 5.2.2. These valves were designed with a corrosion allowance for 40 years of operation. The evaluation of the effects of corrosion on these PORVs is based on a predicted corrosion rate for the plant lifetime and is therefore a TLAA. The applicant stated that because the valves were installed more than 20 years into the current license, the corrosion rate and allowance remain applicable for the period of extended operation.

4.7.2.1.2 Staff Evaluation The staff concurs with the applicant that the 40-year PORVs corrosion rate and the corrosion allowance remain valid since the remaining life of the PORVs will exceed the period of extended operation. The staff concludes that, in accordance with 10 CFR 54.21(c)(1)(i), the applicant has demonstrated that the margin against excessive corrosion which is specified for 40-year plant life will be maintained during the period of extended operation.

In RAI 4.7.2.1, the staff requested that the applicant include a section in the supplement to the UFSARs regarding the corrosion allowance TLAA for PORVs. In its response dated December 17, 2003, the applicant provided supplemental Section A.3.5.2.3 and committed to include it in the UFSARs supplements. Subsequently, the staff reviewed page A-57 of the applicants submittal dated March 5, 2004 and found that Section A.3.5.2.1 of the UFSAR Supplement

4-46 clarified that the corrosion allowance for the Quad Cities, Unit 2 PORVs were evaluated and found to be valid for the period of extended operation. The staff finds this acceptable.

4.7.2.1.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) that the analyses remain valid for the corrosion allowance for PORV TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the corrosion allowance for PORV TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.2.2 Degradation Rates of Inaccessible Exterior Drywell Plate Surfaces 4.7.2.2.1 Summary of Technical Information in the Application In Section 4.7.2.2 of the LRA, the applicant stated that the Mark I containment design (all four units) includes an inaccessible sand pocket around the drywell. The applicant recognized that the potential for degradation of the containment exists because of conditions that allow the introduction of water into the annulus (expansion gap) between the containment and the primary containment shield wall. Water can be introduced from leakage of the refuel cavity past the refueling bellows drain line expansion joints during refueling or because of the introduction of water at other drywell penetrations. This water migrates to the sand pocket under the bottom elevation of the containment and then passes through the sand pocket drain lines. The applicant explains that if the drain lines become clogged, the water remains in the sand pocket and creates an environment that may be corrosive to the containment steel plates.

In response to GL 87-05, Potential Degradation of Mark I Drywell, Dresden and Quad Cities projected corrosion rates for the steel drywell plates in this area and determined that the wall thickness was sufficient for the remainder of the 40-year license period. The applicants evaluation of the remaining life of the drywell steel thickness based on a specified corrosion rate is a TLAA.

Also in response to GL 87-05, the applicant stated that it performed an inspection of the drain lines on all four units to detect leakage in the pocket region. As a result of these actions, the applicant determined that the sand pocket drains were clogged on Dresden Unit 3 and performed an evaluation of actual plate thickness on this unit. The applicant also stated that the design of the Dresden Unit 3 containment vessel is such that margin exists between the required shell thickness and the actual thickness of the steel plate provided. A reevaluation of the required shell thickness in the region of the sand pocket was performed based on loads and data compatible with the original certified containment vessel stress report by Chicago Bridge &

Iron Company. It was determined that the thickness of the plates in the sand pocket region may be reduced to approximately 0.25 in. below the nominal and still be within ASME Code allowable stress limits.

Actual UT thickness measurements were made of the Dresden Unit 3 drywell steel plate at the sand pocket level. All thickness measurements were on the high side (above the nominal

4-47 1.0625 in.). The measurements were taken during the 18th year of operation for Dresden Unit

3. In response to GL 87-05, the applicant stated that it made conservative estimates of corrosion rates that might occur. Starting with the minimum as-found steel plate thickness of 1.08 in. and assuming a corrosion rate of 0.01 in. per year, the applicant indicated that 27 years of service would remain before the effects of corrosion on stresses would become significant.

The corrosion rate was based upon a worst-case rate of 10 mils per year (mil/y) for fresh river water.

The final response to GL 87-05 indicated that the amount of moisture found in the sand pocket drains at Quad Cities was negligible in comparison to that at Dresden Unit 3 and that it was not expected that any corrosion occurred on either unit. The final response also indicated that there was no reason to expect adverse thickness of the drywell liner on Dresden Unit 2.

However, the Dresden Unit 3 plate thickness estimates were used to bound Dresden Unit 2 and both units at Quad Cities. The conservatively analyzed years of service life (18 + 27 = 45) would not be sufficient for the extended license period.

Based on the above findings, the applicant used the provision of 10 CFR 54.21(c)(1)(ii) (i.e., the projected analysis will be valid during and to the end of the extended period of operation), to state the following:

The corrosion rate assumptions used in the calculation will be confirmed by a UT inspection prior to the period of extended operation. The inspection will be performed at Dresden and the results will be used to revise the corrosion calculation and validate that an acceptable wall thickness will remain to the end of the 60 year license operating period.

4.7.2.2.2 Staff Evaluation The staff has identified the following three principal concerns regarding the applicants analysis:

failure to address the preventive measures that would alleviate the root cause of the problem (i.e., preventing water leakage from refueling canal) sole dependence on Dresden Unit 3 examinations for monitoring the condition of the drywell shell in the other units no consideration of potential drywell shell corrosion in the inaccessible portion of the straight portion of the drywell In Section 4.7.2.2.1 of this SER, the applicant recognized the root cause and stated the following:

The potential for degradation of the containment exists due to conditions that allow the introduction of water into the annulus (expansion gap) between the containment and the primary containment shield wall. Water can be introduced due to leakage of the refuel cavity past the refueling bellows drain line expansion joints during refueling or due to the introduction of water at other drywell penetrations. This water migrates to the sand pocket under the bottom elevation of the containment and then passes through the sand pocket drain lines.

The applicant explained that if the drain lines become clogged, the water remains in the sand pocket and creates an environment that may be corrosive to the containment steel plates.

4-48 In RAI 2.4-3, the staff requested information regarding the aging management of refueling seals through which the water has leaked in the past and is likely to leak during the extended period of operation. The applicant provided the following partial response:

The refueling seals (the drywell-to-reactor building refueling) seal and the reactor pressure vessel (RPV)-to-drywell refueling are not safety-related, and they are not relied upon to remain functional during design basis events to ensure (i) the integrity of the reactor coolant pressure boundary, (ii) the capability to shutdown the reactor and maintain it in a safe shutdown condition, or (iii) the capability to prevent or mitigate potential offsite exposures comparable to those referred to in 10 CFR 50.34(a)(1), 50.67(b)(2), or 100.11. Thus, the refueling seals are not brought into scope of license renewal by 10 CFR 54.4(a)(1).

The staff reviewed the applicants response and identified the following potential concerns resulting from the water leakage through the drywell-to-refueling (DR) seal:

corrosion of the steel shell between the straight portion of the drywell and the containment shield wall corrosion of the steel shell in the sand pocket areas A small amount of water leakage may be absorbed by the insulation in the annulus space and may not affect the sand cushion areas. However, this moisture accumulation could corrode the drywell shell plate. The applicants TLAA addresses only the corrosion of the shell in the sand pocket areas. The steel shell in the annulus area is inaccessible for direct inspection.

Based on industry-wide and plant-specific operating experience with the performance of the DR seal, the staff believes that managing the performance of this seal will minimize the potential for corrosion in both these areas. Accordingly, to complete its review, the staff requested the applicant to provide the following supplemental information:

the methodology for managing the potential corrosion of the drywell shell in the inaccessible annulus areas.

justification for not managing the DR seal as a preventive measure against corrosion of the drywell shell plate.

In its response, the applicant emphasized that no operating experience records were found to support an assumption that water had actually accumulated in this area and that related wall thinning had actually occurred. However, the applicant also agreed that, Potential corrosion that could lead to wall thinning of the annulus areas of the drywell shell can be postulated based upon an assumption of moisture accumulation in the annulus space. The applicant also discussed the industry experience related to corrosion of the cylindrical portion of the drywell, and proposed the following monitoring program:

Exelon proposes that a monitoring program be instituted for the Dresden Unit 3 inaccessible annulus areas to ensure that potential corrosion has not occurred. As previously described, Dresden Unit 3 is considered the limiting case for potential drywell corrosion among the four Dresden and Quad Cities units.

The program will consist of inspection of a sample of locations in the cylindrical and upper spherical areas of the drywell, using ultrasonic measurements of the drywell shell thickness made

4-49 from accessible areas of the drywell interior. A minimum of four sample locations will be selected in each 90-degree quadrant of the drywell. At least one of the four sample locations will be performed in the spherical portion of the drywell below the annulus region in proximity to the location of the 1986 drywell liner fire.

A baseline inspection will be performed prior to the period of extended operation. A follow-up inspection consisting partly of the same locations and partly of variable locations will be based on the operating experience records, Exelon believes no detrimental corrosion has or will occur on the exterior of the cylindrical portion of the drywell shell. Regardless, Exelon proposes that a monitoring program be instituted for the Dresden Unit 3 inaccessible annulus areas to ensure that potential corrosion does not occur. As previously described, Dresden Unit 3 is considered the limiting case for potential drywell corrosion among the four Dresden and Quad Cities units.

As the same location will allow trending of wall thickness data, variable locations will allow flexibility to ensure that wall thickness integrity is maintained during the period of extended operation. The follow-up inspection will be used to determine whether any corrosion is occurring and that any observed corrosion rate will not threaten drywell integrity during the extended 60-year plant life.

The results of the inspections will undergo an engineering evaluation to determine if further follow-up inspections are warranted, or if more locations in the accessible drywell liner interior should be monitored.

These inspections will be added to LR program B.1.26, ASME Section XI, Subsection IWE, for Dresden 3.

This is part of Commitment #26 of Appendix A of this SER. The staff finds the proposed monitoring program acceptable, as it will ensure that any corrosion occurring in the inaccessible areas of the cylindrical and spherical portions of the drywell will be monitored and trended, and corrective actions will be taken to ensure the integrity of the drywell during the extended period of operation.

Regarding the second concern, the staff reviewed the applicants detailed response to RAI 3.5-5, which includes a table showing the UT measurements of the Dresden Unit 3 shell during the inspections performed in 1988, 1997, 1998, 2000, and 2002. The staff finds the applicants justification for not performing periodic UT examination for the remaining three units acceptable, provided the applicant has in place a surveillance program to periodically verify the performance of sand-pocket drains in all units. The measured shell wall thickness in Dresden Unit 3 (based on the table) is well within the 10 percent allowable reduction of Subsection IWE of the ASME Code. However, the 0.25 in. allowable corrosion limit (about 25 percent of the shell thickness), which has been established as the long-term acceptance criterion, appears to be excessive. To address this concern regarding the corrosion of the drywell steel plate and to complete its review, the staff requested the following supplemental information:

clarification of whether the performance of sand drains is (and will be) monitored at every outage on all four units.

clarification of where the engineering evaluation performed to set the 0.25 in. (below the nominal thickness of the shell) criterion incorporates the effects of discontinuities and stress concentration under all load combinations. Provide a summary of the calculations performed.

the procedures used for re-coating any corroded steel areas.

4-50 In its response, the applicant provided the following information about the drywell liner surveillance program:

Dresden and Quad Cities conduct a surveillance of the drywell liner drains once per operating cycle to make sure that there is no leakage from the drywell liner sand pocket drain lines on all four units. The surveillance is conducted during refueling operations when the refueling cavity is flooded and the potential for water leakage exists. These surveillances are conducted under the direction of Quad Cities Technical Specification 0820-11, Surveillance of Dryer Separator Pool, Spent Fuel Pool, and Drywell Liner Drains, and Dresden Quad Technical Specifications 1600-06, Drywell Liner Leakage Inspection. Both site procedures have been credited for aging management and are included under aging management program B.1.26, ASME Code,Section XI, Subsection IWE.

In response to a follow-up question from the staff, the applicant also stated that the Dresden inspection procedure currently includes steps to ensure that the sand pocket drains are clear. A similar requirement will be added to the Quad Cities procedure. This is part of Commitment #26 of Appendix A of this SER.

As stated in the response to RAI 3.5-5(b), the reference to a minimum required plate thickness of 0.25 in. below nominal at the sand pocket region is based on information contained in Section 6.2.1.2.1.2 of the Dresden UFSAR. The calculation took the existing design loads and load combinations from the original certified containment vessel stress report generated by Chicago Bridge & Iron Company, and used those cases to calculate the minimum required thickness. Normal operating, refueling, and accident loads were included in the calculation using ASME Code Case N-284 stress allowable limits.

As stated in the applicants response to RAI 3.5-5(b), initial thickness measurements supported the conclusion that significant corrosion was not occurring in the drywell steel plate at the sand pocket region. All measurements to date at Dresden Unit 3 have remained within the 10 percent limit below nominal of Subsection IWE of the ASME Code (most measurements are still above nominal wall thickness). Evaluations performed following the fire in 1986 concluded that the zinc chromate primer coat on the outer shell surface was intact following the fire, and that no corrosion had taken place. No recoating activities have been performed on the drywell exterior. Any recoating activities performed on the drywell interior are performed using proper procedures. No recoating activities performed to date are related to corrosion of the sand pocket region.

With the procedural commitment to inspect the sand pocket drains in all four units of Dresden and Quad Cities, and to explicitly describe the procedure for evaluating the corroded areas, when found, the staff finds the applicants response acceptable. This is Commitment #50 in Appendix A of this SER.

A review of the UT measurement table provided as part of the applicants response to RAI 3.5-5 indicates that within two digits after the decimal points, there is no consistency in the measured results. Some measurements taken at the same locations (e.g., locations 157.5.1.1A and 202.5.1.1A) show an increase in thicknesses in subsequent years, after the 1988 measurements. To complete its review, the staff requested the following supplemental information regarding these measurements:

4-51 Provide the basis for selecting the locations where the UT measurements are taken. State whether the measurement locations indicated in the table have shown visual evidence of corrosion, or if they are in close proximity to the corroded areas.

One location, 337.5.1.2B, shows gradually decreasing thickness, with a total reduction in thickness of 0.18 in. Although the minimum thickness is still close to the nominal thickness, indicate if this reduction implies a 0.18 in corrosion of the shell at that location between 1988 and 2002.

State whether the nominal thickness of 1.0625 in. is the same for the drywell shells in all four units.

Provide the permissible tolerances of the nominal shell thickness in the as-delivered condition. Indicate if the records of the actual as-delivered thicknesses are available.

Provide a discussion of the accuracy of the UT measurements.

In response, the applicant provided the following additional information for evaluating the existing corrosion in the Dresden Unit 3 sand pocket areas:

The UT measurements are taken at various locations around the floor of the Dresden Unit 3 drywell in the sand pocket region. This area was chosen in 1988 as the most likely area to experience potential corrosion resulting from the presence of moisture (Reference 1). The drywell floor is a poured concrete slab that fills the bottom portion of the steel enclosure to an elevation of 502-4 (refer to elevation view of the containment sand pocket, Figure 6.2-9 of Dresden UFSAR). The slab was divided into sectors and measurement locations were selected at random within each sector. The process used a standard statistical sampling basis. Certain portions of the floor slab are inaccessible because of equipment mounted on the floor. These locations are therefore not part of the measurement population. This sampling approach was described in detail in the August 1988 letter from ComEd (W.E.

Morgan) to the NRC.

The shell was made accessible for measurement by drilling 2.5 in. core holes in the concrete at 22 accessible measurement locations. The holes range in depth from 1 to 3 feet, and are capped during normal plant operation. The UT probe is inserted into each hole to provide the thickness measurement. No visual inspection of the shell is possible because of the slab inside the steel liner and the sand pocket outside the steel liner.

Evaluations performed following the fire in 1986 concluded that the zinc chromate primer coat on the outer shell surface was intact following the fire, and that no corrosion had taken place.

The point in question had an abnormally high reading in the 1988 measurements. In the Reference 1 submission, it was speculated that one non-standard plate section could have been substituted in the drywell shell, resulting in thickness measurements exceeding the nominal tolerance on the high side at some locations. However the recent measurements from 1997-2002 have been relatively consistent at all of these locations and do not provide indication of progressive wall thinning.

4-52 The drywell shell thickness varies throughout the structure. The nominal thickness of 1.0625 in. in the sand pocket region applies to Dresden, and bounds the Quad Cities nominal thickness of 1.25 in. in the same area.

No as-delivered thickness measurements are available. Based upon normal mill tolerances for 1-1/16 in. plate a range from 1.0525 in. to 1.1755 in. could be expected.

The initial UT measurements in 1988 were performed with an instrument calibrated to 0.020 in., using a carbon steel standard ranging in thickness from 0.25 in. to 2.0 in. (as described in the August 1988 letter). The instrumentation in use for current UT measurements should attain comparable accuracy. Per current Exelon ultrasonic measurement procedures, UT instrumentation for material thickness measurements is calibrated to an accuracy of 2 percent, which is the same accuracy that was used in 1988.

This response provides the process used in evaluating the findings of corrosion in the sand pocket areas of Dresden Unit 3. The staff has reviewed this response and finds the sampling and measurement calibration process used in the evaluation to be acceptable, as the TLAA based on these initial findings will be continued through the period of extended operation, and will ensure the integrity of the containment drywells at Dresden and Quad Cities.

In Section A.3.5.2.1 (UFSAR Supplement) of the LRA, the applicant summarized this TLAA and reiterated that the calculations will be revised for the realistic environment and for a full 60-year design life in accordance with 10 CFR 50.54.21(c)(1)(ii). The staff finds the UFSAR Supplement adequate.

4.7.2.2.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(ii), that the analyses have been projected to the end of the period of extended operation for the degradation rates of inaccessible exterior drywell plate surfaces TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the degradation rates of inaccessible exterior drywell plate surfaces TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.2.3 Galvanic Corrosion in the Containment Shell and Attached Piping Components due to Stainless Steel Emergency Core Cooling System Suction Strainers 4.7.2.3.1 Summary of Technical Information in the Application In Section 4.2.7.3 of the LRA, the applicant stated that suction strainers at Dresden and Quad Cities were replaced with larger stainless steel strainers, to address potential plugging and unacceptable head loss concerns. This modification created direct contact between the carbon steel support flanges and the stainless steel strainer components through uncoated bolt holes in the carbon steel flanges, resulting in galvanic corrosion effects. The evaluation of the effects of galvanic corrosion on the ECCS suction strainer flanges is based on a predicted corrosion rate for the plant lifetime and is therefore a TLAA. The calculation of the corrosion effects on

4-53 the support flanges assumes a corrosion rate of 4 mil/y and a design life of 33 years, which is not sufficient to encompass the entire period of extended operation.

4.7.2.3.2 Staff Evaluation In RAI 4.7.2.3(a), the staff requested that the applicant describe (1) the nature of the TLAA calculations and the basis for the assumed corrosion rate of 4 mil/y for 33 years used in the calculation, and to (2) substantiate that this is a bounding corrosion rate for all foreseeable conditions in both plants, including any credible nonstandard water chemistry conditions.

The applicant stated that calculations were performed to evaluate and qualify the bolted flange connections between the ECCS suction strainers and the associated torus penetration nozzles.

During an earlier NRC review of the modifications associated with the replacement of the suction strainers, the NRC identified a concern that the calculations did not sufficiently account for the effects of galvanic corrosion. These calculations were revised to include a corrosion allowance in the determination of the stresses at the bolt circles in the existing carbon steel flanges, in accordance with ASME Code,Section III, Subsection NC, 1977 Edition with Summer 1997 Addenda.

The corrosion rate (4 mil/y) was obtained from Uhligs Corrosion Handbook, 2nd Edition, a standard industrial reference source on corrosion. The revised calculation did not include any specific consideration of nonstandard water chemistry conditions. However, the subject strainers are located within the suppression chambers of the respective units. Water quality is maintained within strict limits and sampled quarterly in accordance with plant procedures. It is sampled frequently enough to allow prompt identification and correction of any nonstandard water chemistry conditions capable of increasing corrosion of the flanges. The staff finds the applicants justification acceptable because it conforms with standard industry practice.

In RAI 4.7.2.3(b), the applicant was requested to show how the location of a single ultrasonic inspection, to confirm the assumptions used in the corrosion rate calculations will be selected, and how it will represent the most aggressive corrosion conditions in both plants. The applicant stated that UT inspections will be performed on a randomly selected flange in each Dresden unit. Initial thickness measurements will be made at two to four adjacent bolt locations on the flange. Separate thickness measurements at the same locations will be made in a subsequent outage to establish the actual corrosion rate. The applicant also stated that this method of selection is acceptable to provide assurance that the results are representative of the most aggressive corrosion conditions at both plants because (1) the water chemistries are similar at each site and are required to be maintained within limits established by the same procedures, (2) the strainers at both plants were installed in approximately the same time frame (1997/1998), and (3) the strainer flange configurations are similar at both plants. The staff finds that the applicant has provided a reasonable procedure for determining corrosion rates, and has provided adequate justification to show that the corrosion rates will be similar for all ECCS strainer flanges at both plants.

In RAI 4.7.2.3(c), the applicant was requested to state the corrective measures that will be taken in the event that the revised galvanic corrosion calculation indicates that an unacceptable reduced flange wall thickness will be reached before the end of the period of extended operation. The applicant stated that in this event, the Exelon Corrective Action Program will be used to develop appropriate corrective actions, among which are the inspection of additional

4-54 flanges, establishment of root causes and corrective actions including periodic UT inspections of the affected flanges, and implementation of possible modifications and/or replacement of the affected flanges and/or interfacing components. The staff finds this acceptable because the applicant has made a commitment to take appropriate measures in the event that galvanic corrosion is determined to potentially affect the structural integrity of the strainer flanges and suppression chamber attachments. This is Commitment #50 in Appendix A of this SER.

The staff has also reviewed Section A.3.5.2.2 in the supplement to the Dresden and Quad Cities UFSARs. The section does not reflect the applicants commitment to take appropriate measures in the event that galvanic corrosion is determined to potentially affect the structural integrity of the strainer flanges and suppression chamber attachments. In RAI A.3.5.2.2, the applicant was requested to include the commitment to take the appropriate measures in the supplement to the UFSARs. The applicant stated in its response to the RAI, dated December 17, 2003, that a statement will be added to the UFSAR section stating that in the event that the measured galvanic corrosion rate will not ensure acceptable thickness to the end of the 60-year licensed operating period, appropriate corrective action will be identified and implemented to maintain the structural integrity of the strainer flanges. With this additional information, the staff finds Section A.3.5.2.2 acceptable, because it provides a reasonable summary of the information presented in Section 4.7.2.3 of the LRA.

4.7.2.3.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(ii), that the analyses have been projected to the end of the period of extended operation for the galvanic corrosion in the containment shell and attached piping components caused by stainless steel emergency core cooling system suction strainers TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the galvanic corrosion in the containment shell and attached piping components caused by stainless steel emergency core cooling system suction strainers TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.3 Crack Growth Calculation of a Postulated Flaw in the Heat Affected Zone of an Arc Strike in the Suppression Chamber Shell 4.7.3.1 Summary of Technical Information in the Application The applicant reported in the LRA that arc strikes found during 1990 in the Dresden and Quad Cities suppression chamber walls were evaluated using a common analysis. The evaluation included crack growth calculations that assumed 850 load cycles resulting from SRV and other operations over 40 years of plant life. A further evaluation in 1997 determined that the depth of the arc strike at Dresden was not sufficient to warrant any final repairs. Assuming an operating limit of 850 SRV load cycles, the applicant indicated that no further action was warranted. The applicant performed an ultrasonic measurement at Quad Cities that it claimed validated that no flaw existed in the heat-affected zone of the original arc strike. The applicant performed an evaluation and determined that further repairs or inspections were not warranted.

4-55 The LRA noted that the number of SRV actuations used in the Quad Cities containment analysis is 550 for 40 years compared to the 850 actuations assumed for the flaw evaluation.

The applicant stated that the expected number of SRV actuations from the year that the flaw was repaired (1990) to the end of the extended operating period in 2032 would be at most (550/40) x (2032-1990) = 577.5, which is less than the 850 actuations assumed for flaw evaluation. Therefore, the applicant indicated that the evaluation remains valid for Quad Cities for the extended period of operation.

The LRA noted that the number of SRV actuations used in the Dresden containment analysis is 300 for 40 years compared to the 850 actuations assumed for the flaw evaluation. The current license for Dresden Unit 3 will expire in 2011, and the Dresden flaw was repaired in 1991. The applicant stated that the expected number of SRV actuations from the year the flaw was repaired to the end of the extended operating period in 2031 would be at the most (300/40) x (2031-1991) = 300, which is less than the 850 actuations assumed for the flaw evaluation.

Therefore, the applicant indicated that the evaluation remains valid for Dresden for the extended period of operation.

4.7.3.2 Staff Evaluation Per 10 CFR 54.21(c)(1), applicants for license renewal must demonstrate that TLAAs have been projected through the end of the period of extended operation and remain valid for the period of extended operation. Alternatively, the applicant may demonstrate that the effects of aging that are applicable to the components evaluated by the TLAAs will be managed during the period of extended operation. The applicant stated that the crack growth calculations for arc strikes will project through the end of the period of extended operation.

The staff reviewed the information provided in the LRA and in a letter dated August 7, 2003, requested that the applicant clarify information provided on this TLAA. In RAI 4.7.3, the applicant was requested to clarify if flaws were actually detected or were only postulated flaws, to clarify if ASME Code Section XI fracture mechanics methodology and acceptance criteria were used for evaluation of crack growth, and to describe any alternative method and acceptance criteria that were used.

In a letter dated October 3, 2003, the applicant responded to the staffs request for additional information. The response to the RAI indicated that the flaws were largely of a postulated nature and were assigned to bound possible damage at the arc strikes. In many cases, further ultrasonic examination indicated that no flaws were detected at the locations of concern. The applicant further responded that fracture mechanics evaluation applied the acceptance criteria of the ASME Code as defined in NUREG-0661, Safety Evaluation Report, Mark 1 Containment Long Term Program, Resolution of Generic Technical Activity. Because the applicant applied the methodology of the ASME Code which has been previously approved by the staff, the staff finds the response acceptable.

4.7.3.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) that the analyses remain valid for the crack growth calculation of a postulated flaw in the heat-affected zone of an arc strike in the suppression chamber shell TLAA. The staff also concludes that the UFSAR Supplement

4-56 contains an appropriate summary description of the crack growth calculation of a postulated flaw in the heat-affected zone of an arc strike in the suppression chamber shell TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.4 Radiation Degradation of Drywell Shell Expansion Gap Polyurethane Foam 4.7.4.1 Summary of Technical Information in the Application The applicant describes the analysis of radiation degradation of the drywell shell expansion gap polyurethane foam in Section 4.7.4 of the LRA. The drywell shell is described as being largely enclosed within the structural and shielding concrete of the reactor building. To accommodate thermal expansion of the drywell, compressible foam was used to form an expansion gap between the concrete and the drywell shell. An analysis performed by the applicant evaluated the increase in external compressive loads on the drywell exterior from additional compression of this foam for normal, refuel, and accident conditions. The effect on this analysis of a postulated increase in the foam stiffness resulting from a radiation dose is a TLAA.

The applicant stated that the polyurethane foam material was chosen for its resistance to the environmental conditions likely to exist during its service life and with characteristics such that the effects of compression during a LOCA resulting in thermal expansion of the drywell would not exceed ASME Code allowable limits. Test results established that there was no detectable change in resilience below 10 8 rads. The original design considered the effects of a 40-year lifetime dose of 2.5 x 10 7 rads on the foam material. Thus, the applicant stated that the resilient characteristics of the polyurethane foam will remain intact during the 40-year design life.

The LRA indicated that a 20-year increase in the design lifetime to 60 years, combined with approved increases in power rating, would conservatively result in a total radiation exposure of 4.2 x 10 7 rads which is less than the 10 8 rads qualified radiation exposure. Therefore, the applicant indicated that material properties will remain within the limits assumed by the original design analysis, in accordance with the aging assumptions assumed by the original design, for the 60-year extended operating period.

4.7.4.2 Staff Evaluation Per 10 CFR 54.21(c)(1), applicants for license renewal must demonstrate that TLAAs for license renewal have been projected through the end of the period of extended operation for their facilities, remain valid for the period of extended operation, or demonstrate that the effects of aging that are applicable to the components evaluated by the TLAAs will be managed during the period of extended operation.

The staff reviewed the information provided in the LRA and in a letter dated August 7, 2003, requested additional information from the applicant related to the test methods used and factors considered in determining the radiation stability of the polyurethane (RAI 4.7.4). In a letter dated October 3, 2003, the applicant responded, noting that the irradiation testing of the polyurethane foam was conducted more than 30 years ago and that additional information on

4-57 the tests is not available. For justification that the tests provide a satisfactory basis for predicting of the acceptable dose within the required uncertainty limits, the applicant relied on the test results showing no predictable loss of resiliency below 108 rads and the predicted end of life exposure to be 4.2 x 10 7 rads. The staff identified that the polyurethane foam radiation limit was previously reviewed by the staff in an SER, and the limit is considered part of the applicants CLB. The staff finds the applicants analysis acceptable because the analysis has been projected through the end of the period of extended operation and remains within the limits established in the applicants CLB.

Based on the review of the LRA and the applicants response to RAI 4.7.4, the staff found that the analysis performed to evaluate the effects of extended life on the radiation degradation of the drywell shell expansion gap polyurethane foam is appropriate and provides a basis for concluding that the safety margins established and maintained during the current operating period will be maintained during the extended period of operation.

4.7.4.3 Conclusions On the basis of its review, the staff concludes that the applicant has provided an acceptable demonstration, pursuant to 10 CFR 54.21(c)(1)(i) that the analyses remain valid for the radiation degradation of drywell shell expansion gap polyurethane foam TLAA. The staff also concludes that the UFSAR Supplement contains an appropriate summary description of the radiation degradation of drywell shell expansion gap polyurethane foam TLAA evaluation for the period of extended operation, as required by 10 CFR 54.21(d). Therefore, the staff has reasonable assurance that the safety margins established and maintained during the current operating term will be maintained during the period of extended operation, as required by 10 CFR 54.21(c)(1).

4.7.5 High-Energy Line Break Postulation Based on Fatigue Cumulative Usage Factor This issue is included only because it is listed as a possible TLAA in NUREG-1800. Neither the Dresden nor the Quad Cities postulated break locations are based on a fatigue usage factor criterion, nor are any break locations based on any other evaluation of fatigue effects. This is not a TLAA for either Dresden or Quad Cities.

4.8 Conclusion for Time-Limited Aging Analyses The staff has reviewed the information in Section 4 of the LRA. On the basis of its review, the staff concludes that the applicant has provided an adequate list of TLAAs, as defined in 10 CFR 54.3. Further, the staff concludes that the applicant has demonstrated that the TLAAs (1) will remain valid for the period of extended operation, as required by 10 CFR 54.21(c)(1)(i), (2) have been projected to the end of the period of extended operation, as required by 10 CFR 54.21(c)(1)(ii), or (3) the aging effects will be adequately managed for the period of extended operation, as required by 10 CFR 54.21(c)(1)(iii). In addition, the staff concludes that there are no plant-specific exemptions in effect that are based on TLAAs, as required by 10 CFR 54.21(c)(2). On this basis, the staff has reasonable assurance that the aging effects associated with the structures and components subject to TLAAs will perform their intended functions in accordance with the CLB during the period of extended operation, as required by 10 CFR 54.21(a)(3).

ML042050503

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