ML110830978
| ML110830978 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/14/2011 |
| From: | Daily J License Renewal Projects Branch 1 |
| To: | Gerry Powell South Texas |
| Daily, J NRR/DLR, 415-3873 | |
| References | |
| TAC ME4936, TAC ME4937 | |
| Download: ML110830978 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 14, 2011 Mr. G. T. Powell Vice President, Technical Support and Oversight STP Nuclear Operating Company P.O. Box 289 VVadsworth, TX 77483
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEVV OF THE SOUTH TEXAS PROJECT, LICENSE RENEVVAL APPLICATION - FIRE PROTECTION AND COMPONENT INTEGRITY (TAC NOS. ME4936 AND ME4937)
Dear Mr. Powell:
By letter dated October 25, 2010, STP Nuclear Operating Company submitted an application pursuant to Title 10 of the Code of Federal Regulation Part 54 to renew the operating licenses NPF-76 and NPF-80 for South Texas Project Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with Mr. Arden Aldridge, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at (301) 415-3873 or bye-mail john.daily@nrc,gov.
Sincerely, John Daily. Sr. Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation DocKet Nos. 50-498 and 50-499
Enclosure:
Requests for Additional Information cc w/encl: Listserv
SOUTH TEXAS PROJECT LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION RAI2.3.3.17-1
Background:
For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E:
Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.
Issue:
The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of Federal Regulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.
Request:
The following LRA boundary drawings shows the following fire protection systems/components as out of scope (Le., not colored in green):
LRA Drawing Systems/Components Location LR-STP-FP-Fire water suppression systems associated with Fire Protection 7Q271 F00046 transformers BOP 101 and 102 Loop LR-STP-FP-Fire water suppression system in the Lighting Fire Protection 7Q271 F00046 Diesel Generator Building Loop LR-STP-FP-Fire water suppression systems associated with Fire Protection 7Q272F00046 transformers BOP 201 and 202 Loop LR-STP-FP-Several fire water suppression systems associated B7 and 08 7Q272F00046 with various buildings, (e.g., Bldg. 15, Bldg. 27, Bldg. 33, Bldg. 45, Bldg. 50, Bldg. 52, and Bldg. 71)
The staff requests that the applicant verify whether the fire protection systems/components listed above are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21 (a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.
ENCLOSURE
- 2 RAI 2.3.3.17-2
Background:
For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E:
Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.
Issue:
The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of Federal Regulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.
Request:
Section 9.5.1.2.1, Fire Protection Water Supply System," of the UFSAR on page 9.5-5, states that the water supply to refill the fire water storage tanks is normally provided from the Fresh Water System, which takes suction from a settling basin. This section also states that in the event of a failure in this system, the tank is refilled directly from the site well water system. LRA Section 2.3.3.17 discusses requirements for the fire water supply system but does not mention site well water pumps and associated components.
The staff notes that LRA boundary drawing LR-STP-FP-7Q270F00006 shows the site well water system and its components as out of scope (Le., not colored in green). The staff requests that the applicant verify whether the site well water pumps and associated components to fire water storage tanks are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21 (a)(1). The staff requested that, if these components are excluded from the scope of license renewal and are not subject to an AMR, the applicant provide justification for the exclusion.
RAI 2.3.3.17-3
Background:
For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E:
Safety Evaluation Report" NUREG-0781, dated April 1986, and its Supplements.
- 3 Issue:
The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAI) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of Federal Regulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.
Request:
Tables 2.3.3-17 and 3.3.2-17 of the LRA do not include the following fire protection components:
fire hose stations, fire hose connections, and hose racks floor drains for fire water dikes and curbs for oil spill confinement components in reactor coolant pump oil collection system The staff requests that the applicant verify whether the fire protection components listed above are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21 (a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.
RAI2.4-1 Backqround:
For South Texas Project (STP), Units 1 and 2, the staff reviewed the license renewal application (LRA); drawings; updated final safety analysis report (UFSAR), Section 9.5.1, and Fire Protection Evaluation and Comparison to Branch Technical Position (BTP), Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," May 1, 1976. The staff also reviewed the following fire protection documents cited in the current licensing basis listed in the STP, Units 1 and 2, "Operating License Conditions 2.E:
Safety EValuation Report" NUREG-0781, dated April 1986, and its Supplements.
Issue:
The staff has identified that instances of fire protection systems and/or components as noted in this request for additional information (RAt) have been excluded from the scope of license renewal and/or an aging management review (AMR). These systems/components that were not included in the license renewal boundaries appear to have fire protection intended functions required for compliance with Title 10 of the Code of Federal Regulations (CFR) 50.48, "Fire Protection," as stated in 10 CFR 54.4. Therefore, in order to complete our review, the staff requires a response to the following request.
Request:
Section 2.4 of the LRA does not include the following fire barrier components:
Table 2.4-1, fire barrier seals
-4 Table 2.4-4, concrete elements, concrete wall (masonry walls)
Table 2.4-8, fire barrier doors Table 2.4-9, fire barrier coatings The staff requests that the applicant verify whether the above fire barrier components are in the scope of license renewal in accordance with 10 CFR 54.4(a) and subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.
Section 4.3.2.1 Reactor Pressure Vessel, Nozzles, Head and Studs RAI 4.3.2.1-1
Background:
Section 4.3.2.1, page 4.3-11, states that the STP, Units 1 and 2, reactor pressure vessel (RPV) heads were replaced in the Fall of 2009 and the Spring of 2010, respectively.
Issue:
The staff notes that some replacement RPV heads have detected fabrication defects prior to installation in nuclear plants. Therefore, the staff is seeking information regarding the nondestructive examination data in the replacement RPV heads at STP units.
Request:
(1) Discuss any recordable indication(s) detected in the new RPV heads in both units.
(2) Discuss how the indications, if any, will be monitored to the end of 60 years.
(3) Discuss measures that have been taken and the design features in the replacement RPV head to minimize the degradation in control rod drive mechanism penetration nozzles.
RAI4.3.2.1-2
Background:
Section 4.3.2.1, page 4.3-11, second paragraph, states that the closure flanges, studs and nuts are designed for 107 tightening cycles. A cycle consists of the reactor head removal and re installation. It also states that the number of reactor head removals and re-installations to support refueling operations in 60 years is estimated to be 80-cycles.
Issue:
In the stress analysis of the closure flanges, studs and nuts, transient cycles are an input parameter to determine the fatigue usage factor. It is not clear which cycles described above are used in the fatigue usage factor calculation.
Request Clarify how many cycles (80, 107, or 187) were used in calculating the fatigue usage factor for closure flanges, studs and nuts.
- 5 RAI 4.3.2.1-3
Background:
Section 4.3.2.1, page 4.3-11, second paragraph, states that the maximum usage factor based on the designed number of transient cycles in the closure flanges is 0.089.
Issue:
The staff is not. clear as to why the fatigue usage factor for the RPV head closure flange is an order of magnitude lower than that of the stud inserts and studs. The staff also seeks clarification on the terminology of the "maximum usage factor."
Request:
(1) Discuss why the flanges have such a low usage factor when compared to the stud inserts (0.8852) and studs (0.3372).
(2) Discuss whether the maximum usage factor is the same as the cumulative usage factor (CUF) as specified in the ASME Code,Section III.
RAI 4.3.2.1-4
Background:
Section 4.3.2.1, page 4.3-12, first paragraph, states that the corrosion analyses, fatigue analyses, and the ASME Section XI 48-year flaw growth analysis for the bottom mounted instrumentation nozzles and lower head repairs are valid for the period of extended operation.
Therefore these time-limited aging analyses (TLAAs) are dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).
Issue:
The applicant stated without detailed discussion that the corrosion analyses, fatigue analyses and flaw growth analysis for the bottom mounted nozzles and lower head repairs are valid for the period of extended operation. The applicant needs to provide a detailed explanation as to why these analyses are valid for the period of extended operation.
Request:
Explain in detail how these analyses are valid for the period of extended operation per 10 CFR 54.21 (c)(1 )(i).
RAI 4.3.2.1-5
Background:
Section 4.3.2.1, page 4.3-12, second paragraph, states that the fatigue analyses for the replacement reactor vessel closure head are valid for the period of extended operation.
Therefore, this TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).
- 6 Issue:
The applicant did not explain why the fatigue analyses for the replacement reactor vessel closure head are valid for the period of extended operation. Also, the staff is not clear whether the "fatigue analyses" imply "fatigue usage factor calculations" or "fatigue crack growth calculations. "
Request:
(1) Clarify how the fatigue analyses for the replacement reactor vessel head are valid for the period of extended operation per 10 CFR 54.21(c)(1)(i).
(2) Confirm that the "fatigue analyses" referred to above are the fatigue usage factor calculations, not the fatigue crack growth calculations.
RAI4.3.2.1-6
Background:
Section 4.3.2.1, page 4.3-13, Table 4.3-3, Footnote (2) states that "". The 40-year design basis number of events should be sufficient for 60 years of operation; therefore the calculated 40-year usage factors will not be exceeded... "
Issue:
The wording "should be" causes the aforementioned statement to be ambiguous.
Request:
Submit the analysis to demonstrate that the 40-year design basis number of events is sufficient for 60 years.
RAI4.3.2.1-7
Background:
Section 4.3.2.1, page 4.3-13, Table 4.3-3, Footnote (3) states that 60-year projections are performed by multiplying the design basis CUFs by 1.5, which is equal to 60 years/40 years, unless otherwise noted. The applicant applied a factor 1.5 to the usage factors calculated for 40 years to project the usage factors for 60 years. This method assumes that the fatigue usage factor is directly proportional to the transient cycles used in the analysis.
Issue:
The staff is especially concerned about the 60-year usage factors for the closure stud hole inserts and closure studs because their 60-year usage factors are either exceeding the allowable or substantial.
Reguest:
Demonstrate that this linear extrapolation of using a factor 1.5 is appropriate or at a minimum conservative.
- 7 RAI 4.3.2.1-8
Background:
The plates and welds in the RPV and associated nozzles, flanges and supports may contain fabrication defects or service induced flaws.
Issue:
The staff seeks to understand the existing condition of the RPV materials in terms of flaws/indications remaining in service, if any, and to determine how the applicant will manage these flaws/indications during the period of extended operation.
Request:
Identify any flaws/indications that are remaining in service in the RPV, nozzles, flanges, or vessel supports. Discuss how these flawslindications will be managed to the end of 60 years.
Section 4.3.2.4 Pressurizer and Pressurizer Nozzles RAI 4.3.2.4-1
Background:
Section 4.3.2.4, page 4.3-17, first paragraph, states that pressure-retaining and support components of the pressurizer are subject to an ASME Section III fatigue analysis.
Issue:
This analysis has been revised for plant modifications with redefined loads, and for newly identified design basis events not included in the original analyses.
Request:
(1) Describe in detail how the plant modifications affect the loads on the pressurizer and associated components (closures, nozzles, heaters, and support skirts).
(2) Discuss the redefined loads and newly-identified design basis events not included in the original analyses.
(3) Page 4.3-18 discusses the new loads and design basis events but without details. For example, where the cold overpressure mitigation system transient comes from? Where the 6000 pressure fluctuations come from and what is their impact?
RAI 4.3.2.4-2
Background:
Section 4.3.2.4, page 4.3-17, Table 4.3-4 shows that some cumulative fatigue factors exceed the ASME Code,Section III allowable of 1.0.
Issue:
The staff is concerned about how the applicant will manage the aging effects of the components that have projected fatigue usage factor greater than 1.0 during the period of extended operation.
- 8 Request:
Describe in detail how the metal fatigue of reactor coolant pressure boundary aging management program (AMP) will manage those pressurizer components that have been projected to have a CUF exceeding the ASME Code,Section III, allowable of 1.0 at the end of 60 years. For example, discuss how the operator would know when the fatigue usage factor of a particular pressurizer component approaches 1.0, and the actions that would be taken to assess the components.
RAI 4.3.2.4-3
Background:
Section 4.3.2.4, page 4.3-18, first paragraph, states that "... The stress reports evaluated the effect on the pressurizer of 1 0 cold over-pressurization mitigation system activation events. The contribution of these thermal effects to the fatigue usage can be neglected... "
Issue:
The staff is not clear regarding how the applicant performed the stress analyses to show that the 1 0 cold over-pressurization mitigation system activation events can be neglected in the fatigue usage factor calculation.
Request:
(1) Discuss the stress reports.
(2) Explain why the contribution of these thermal effects to the fatigue usage can be neglected.
RAI 4.3.2.4-4
Background:
Section 4.3.2.4, page 4.3-19, first paragraph, states that"... The fatigue crack growth analysis for pressurizer spray, relief, safety, and surge nozzle preemptive overlays is not a TLAA because the crack is not qualified for the life of the plant, but only the inspection intervaL.. Since the crack is not qualified for the life of the plant, but only the inspection interval, the fatigue crack growth analysis is not a TLAA in accordance with 10 CFR 54.3(a), Criterion 3... "
Issue:
The pressurizer spray, relief, safety, and surge nozzles use nickel-based Alloy 82/182 dissimilar metal weld which is susceptible to primary stress-corrosion cracking (PWSCC). The applicant has installed weld overlays on the Alloy 82/182 welds to mitigate the potential for PWSCC. As part of weld overlay installation, the applicant is required to perform crack growth calculations to project the growth of a postulated flaw or an actual detected flaw in the Alloy 82/182 weld. The staff is not clear why the applicant stated that the crack growth analysis is not a TLAA.
Request:
(1) Confirm that every overlaid Alloy 82/182 welds in pressurizer spray, relief, safety and surge nozzles will be inspected every 10 years.
- a. If not, the staff is concerned with the above statement regarding the fatigue crack
- 9 growth analysis not being a TLAA. If an overlaid Alloy 82/182 is not inspected every 10 years, the fatigue crack growth analysis should be analyzed for the remaining life of the plant to ensure that the flaw will not affect the structural integrity of the pipe/weld to the end of 60 years.
- b. In addition, the transient cycles should be monitored to verify that the fatigue crack growth analysis bounds the actual transient cycles. If this is true, then please justify why this cycle verification is not part of a TLAA per 10 CFR 54.21 (c)(1 )(iii).
(2) The staff understands that plant owners perform fatigue crack growth calculations for overlaid Alloy 82/182 welds using 40 years of transient cycles (sometimes 60 years).
Discuss how many years of transient cycles were used in the fatigue crack growth calculation for the overlaid Alloy 82/182 welds for various pressurizer nozzles.
(3) Discuss how the fatigue crack growth analyses for overlaid Alloy 82/182 welds at pressurizer spray, relief, safety and surge nozzles will be monitored to the end of 60 years (Le., the end of the period of extended operation).
RAI 4.3.2.4-5
Background:
Section 4.3.2.4, page 4.3-19, third paragraph, states that "... As shown in Table 4.3-4, the fatigue analyses of the safety and relief nozzles, and the seismic support lugs demonstrate worst-case 40-year usage factors less than 0.4. When multiplied by 1.5 (60/40) to account for the 60-year period of extended operation, these results do not exceed 0.6, providing a large margin to the code acceptance criterion of 1.0... "
Issue:
The applicant used a ratio of 40 years to 60 years to project the fatigue usage factor from 40 years to 60 years. The staff is not clear if this direct proportional projection is acceptable.
The staff also questions a potential discrepancy in the fatigue usage factor value.
Reguest:
(1) The applicant applied a factor 1.5 to the usage factors calculated for 40 years to project the usage factors for 60 years. This method assumes that the usage factor is directly proportional to the transient cycles used in the analysis. Demonstrate that this linear extrapolation of using the ratio of 40 years vs. 60 years (60/40 =1.5) is appropriate or at minimum conservative to project the 60-year usage factors.
(2) Table 4.3-4 of the LRA lists safety and relief nozzles as having a fatigue usage factor of 0.044 for 40 years and 0.066 for 60 years. The aforementioned statement cited the usage factors of 0.4 and 0.6. Clarify the discrepancy on the usage factors between Table 4.3-4 and the above statement.
RAI 4.3.2.4-6
Background:
Section 4.3.2.4, page 4.3-19, last paragraph, states that"... The fatigue analyses of the remaining subcomponents have been found acceptable for a limiting number of transient
- 10 events. The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section 4.3.1 and B3.1 ensures that the numbers of transients actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means... "
Issue:
The staff is not clear on the terminology used in the aforementioned paragraph. In addition, the staff is not clear how the metal fatigue AMP will monitor the transient cycles to verify the fatigue usage factor calculations.
Request:
(1) Identify the "remaining" subcomponents.
(2) Clarify whether the "fatigue analyses" are the fatigue crack growth analyses or cumulative fatigue usage factor analyses.
(3) Describe exactly how the metal fatigue AMP will monitor the transient cycles and track the fatigue usage factor calculations of components in Table 4.3-4.
(4) Describe in detail the "appropriate corrective actions" and "acceptable means" in the above quoted statement.
Section 4.3.2.11.
Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures RAI 4.3.2.11-1
Background:
Section 4.3.2.11, page 4.3-31, states that"... NRC approval of the use of leak-before-break in the reactor coolant system primary loop piping was granted with STP SER, NUREG-0781, Supplement No.2... " Section 4.3.2.1, page 4.3-32, states that "... Westinghouse determined that the conclusions of the previous LBB analysis for the reactor coolant piping, pressurizer surge line, and accumulator lines remain valid after steam generator replacement... "
Issue:
Based on these two statements, it is not clear exactly how many piping systems have been approved for LBB.
Request:
List the piping systems that have been approved for LBB and that are in the scope of the LRA.
RAI 4.3.2.11-2
Background:
Section 4.3.2.11, page 4.3-32, first paragraph, states that "... The Westinghouse LBB analysis for the primary loop cites a study which determined the effects of thermal aging on piping integrity for a material at thermal embrittlement saturation. Therefore, the fracture mechanics
- 11 evaluation is dependent on material properties not plant life and therefore, is not a TLAA by 10 CFR 54.3(a), Criterion 3... "
Issue:
The staff is not clear on why the cast austenitic stainless steel material in the LBB primary loop piping is not a TLAA even though the cast austenitic stainless steel material is susceptible to time-dependent thermal embrittlement degradation. If saturated fracture toughness is used in the LBB analysis, the LBB analysis of the cast stainless steel material could still be considered under 10 CFR 54.21 (c)(i) or (ii). In addition, based on the aforementioned statement, the staff is not clear whether the applicant had used saturated fracture toughness in the LBB analysis.
Request:
(1) Clarify whether the saturated fracture toughness value due to thermal embrittlement was used in the fracture mechanics evaluation for LBB such that the fracture mechanics evaluation is not a TLAA.
(2) If the saturated fracture toughness is used in the fracture mechanics calculation (i.e., the fracture mechanics evaluation is dependent on material properties not plant life), explain why the fracture mechanics evaluation is not a TLAA.
(3) Cite or submit the reference of the study in the Westinghouse LBB analysis in the above quoted statement.
(4) Justify why the reference of the study has demonstrated the effects of thermal aging on cast austenitic stainless steel and that the material properties (i.e., fracture toughness) is at saturated condition.
RAI4.3.2.11-3 Backqround:
Section 4.3.2.11 page 4.3-32, mid-page, states that"... The metal fatigue of the reactor coolant pressure boundary program described in Section 4.3.1 and B3.1 ensures that the numbers of transients remain below the number actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means... "
Issue:
It appears that the applicant intended to state that the metal fatigue AMP will ensure that the numbers of transients used in the analysis remain higher than the actual numbers experience in operation. However, the staff is not clear whether the fatigue crack growth calculation in the LBB application will be part of the analysis that the metal fatigue AMP will be monitoring.
Request:
(1) Discuss whether the metal fatigue AMP specifically identifies all LBB analyses that will be monitored.
(2) Clarify the "appropriate corrective actions" and "other acceptable means" in the above statement.
(3) Clarify the aforementioned statement.
- 12 RAI4.3.2.11-4
Background:
SRP Section 3.6.3 specifies that LBB piping does not experience active degradation. Licensees have used nickel-based Alloy 82/182 weld material in the LBB primary coolant piping. Alloy 821182 welds are susceptible to PWSCC.
Issue:
The staff seeks to have a complete understanding of the Alloy 82/182 welds in LBB piping in the STP units and how the applicant manages PWSCC in the LBB piping during the period of extended operation.
Request:
(1) Identify the LBB pipes that are constructed using Alloy 82/182 filler weld metal which is susceptible to PWSCC.
(2) Identify the LBB pipes with Alloy 82/182 welds mitigated with weld overlays or mechanical stress improvement process.
(3) Identify the LBB pipes with unmitigated Alloy 82/182 welds. For this group of pipes, discuss the measures that will be implemented to minimize the potential for PWSCC.
(4) Was the LBB evaluation updated to account for any Alloy 82/182 welds in the piping system that were mitigated with weld overlay?
RAI 4.3.2.11-5
Background:
Section 4.3.2.4, pages 4.3-17 and 4.3-18, discuss new (redefined) loads and new design basis events due to plant modifications.
Issue:
The applicant discussed plant modifications which may result new pipe loads. The staff is not clear whether the applicant has considered the new loads in the LBB evaluations.
Request:
(1) Discuss whether the fracture mechanics calculations and fatigue crack growth calculations for all LBB piping have been updated to include these new loads and design basis events. If not, justify the validity of all LBB analyses in light of new loads and design basis events.
- 13 Section 4.7.2 In-Service Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years RAI4.7.2-1
Background:
The ASME Code,Section XI, IWA-3000 and IWB-3000, permit flaws to remain in service. Small flaws may remain in service without performing analysis or successive examinations. Larger flaws are required to be analyzed and examined in successive inspection periods.
Issue:
The applicant stated that it has performed a search and did not find any flaws evaluated for the remaining life of the plant other than those discussed elsewhere in the LRA. The staff seeks to understand how the applicant's search was performed. The staff also seeks to understand the flaws that are remaining in service but do not require evaluation. The purpose is to determine the overall condition of the pipes that are covered under the LRA.
Request:
Section 4.7.2, page 4.7-3, states that a search of the CLB [current licensing basis] did not identify any flaws evaluated for the remaining life of the plant other than those discussed elsewhere in this application. (1) Discuss the sources that have been searched to obtain this information. (2) Discuss whether there are recordable indications/flaws that have remained in service in the piping without a flaw evaluation for pipes within the scope of license renewal application. Discuss how these flaws will be monitored to the end of 60 years.
'.. ML110830978
- concurred via email OFFICE LA:DLR PM:RPB1 :DLR BC:RPB1 :DLR PM: RPB1 :DLR NAME IKing JDaily BPham JDaily DATE 3/28/11 3/30/11 4/4/11 4/14/11
Letter to G. T. Powell from J. Daily, Dated April 14, 2011
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SOUTH TEXAS PRO,JECT, LICENSE RENEWAL APPLICATION DISTRIBUTION:
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